Trade study for kWe class space reactors

Science.gov (United States)

Bost, Donald S.

Recent interest by NASA and other government agencies in space reactor power systems with power levels in the 1 to 100 kWe range has prompted a review of earlier space reactor programs, as well as the ongoing SP-100 program, to identify a system that will best fulfill their needs. The candidate reactor types that were reviewed are listed. They are categorized according to the method of heat removal. The five types are: conduction cooled, heat pipe cooled, liquid metal cooled, in-core thermionic and gas cooled. The UZrH moderated reactor coupled with an organic Rankine cycle power conversion system provides an attractive system for multikilowatt, long lived missions. The reactor requires a minimum development because a similar reactor has already flown and the ORC is being developed for use in the Dynamic Isotope Power System (DIPS) and on the Space Station.

Multimegawatt Space Reactor Safety

International Nuclear Information System (INIS)

Stanley, M.L.

1989-01-01

The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

Application of data mining in three-dimensional space time reactor model

International Nuclear Information System (INIS)

Jiang Botao; Zhao Fuyu

2011-01-01

A high-fidelity three-dimensional space time nodal method has been developed to simulate the dynamics of the reactor core for real time simulation. This three-dimensional reactor core mathematical model can be composed of six sub-models, neutron kinetics model, cay heat model, fuel conduction model, thermal hydraulics model, lower plenum model, and core flow distribution model. During simulation of each sub-model some operation data will be produced and lots of valuable, important information reflecting the reactor core operation status could be hidden in, so how to discovery these information becomes the primary mission people concern. Under this background, data mining (DM) is just created and developed to solve this problem, no matter what engineering aspects or business fields. Generally speaking, data mining is a process of finding some useful and interested information from huge data pool. Support Vector Machine (SVM) is a new technique of data mining appeared in recent years, and SVR is a transformed method of SVM which is applied in regression cases. This paper presents only two significant sub-models of three-dimensional reactor core mathematical model, the nodal space time neutron kinetics model and the thermal hydraulics model, based on which the neutron flux and enthalpy distributions of the core are obtained by solving the three-dimensional nodal space time kinetics equations and energy equations for both single and two-phase flows respectively. Moreover, it describes that the three-dimensional reactor core model can also be used to calculate and determine the reactivity effects of the moderator temperature, boron concentration, fuel temperature, coolant void, xenon worth, samarium worth, control element positions (CEAs) and core burnup status. Besides these, the main mathematic theory of SVR is introduced briefly next, on the basis of which SVR is applied to dealing with the data generated by two sample calculation, rod ejection transient and axial

Space Fission Reactor Structural Materials: Choices Past, Present and Future

International Nuclear Information System (INIS)

Busby, Jeremy T.; Leonard, Keith J.

2007-01-01

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

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

Soviet space nuclear reactor incidents - Perception versus reality

Science.gov (United States)

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.

Some thoughts on the commercial use of reactors in space

International Nuclear Information System (INIS)

Buden, D.; Lee, J.

1986-01-01

The purpose of this paper is to illuminate the major regulatory issues associated with commercialization of space nuclear power. Currently, space reactors are government-owned and approved through the Interagency Nuclear Safety Review Panel (INSRP) and the President while commercial reactors are licensed by the Nuclear Regulatory Commission. The commercial use of reactors in space will open a new regime of regulation; that is public intervenors could enter the space licensing process for the first time. The major issue is responsibility for licensing and operations, but related considerations involve controlling special nuclear materials from aborted launches or abandoned platforms, determining final shutdown and disposal tactics, and solving new design issues such as the need for longer life of space reactor power plants

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

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

ZrH reactor lattice spacing (heat transfer considerations)

International Nuclear Information System (INIS)

Felten, L.D.

1970-01-01

Temperature calculations for a 295 element ZrH reactor at fuel element spacings from 0.010'' to 0.065'' showed a very small dependence of reactor temperature on element spacing. It was found that one variation in coolant channel area (2 zones) was sufficient to satisfactorily shape the radial flow profile for the core. (U.S.)

Thermionic reactors for space nuclear power

Science.gov (United States)

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

1985-01-01

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

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

A Sensitivity Study on the Radiation Shield of KSPR Space Reactor

Energy Technology Data Exchange (ETDEWEB)

Cerba, S.; Lee, Hyun Chul; Lim, Hong Sik; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The idea of a space reactor was realised some decades ago and since that time several research activities have been performed into this field. The US National Aeronautics and Space Administration (NASA) has been developing a small fast reactor called as fission power system (FPS) for deep space mission, where highly enriched uranium (HEU) is used as fuel. On the other hand, other researchers have also surveyed a thermal reactor concept with low enriched uranium (LEU) for space applications. One of the main concerns in terms of a space reactor is the total size and the mass of the system including the reactor itself as well as the radiation shield. Since the reactor core is a source of neutrons and gamma photons of various energies, which may cause severe damage on the electronics of the space stations, the questions related to the development of a radiation shield should be address appropriately. The proposal of a radiation shield for a small space reactor is discussed in this paper. The requirements for the radiation shield have been addressed in terms of maximal absorbed doses and neutron flounces during 10 years of operation. In this study a radiation shield design for a small space reactor was investigated. All the presented calculations were performed using the multi-purpose stochastic MCNP code with temperature dependent continuous energy ENDF/B VII.0 neutron and photon cross section libraries. The aim of this study was to design a neutron and gamma shield that can meet the requirements of 250 Gy absorbed during 10 years of reactor operation. The comparison with a fast reactor design showed that high content of {sup 238}U strongly influences the shielding mass. This phenomenon is due to the higher photon production in case of the KSPR design and therefore the use of high {sup 235}U enrichments and the operation in fast neutron spectrum may be more desirable. In case if the KSPR space reactor the best shielding performance was achieved while utilizing a multi

Inform@ed space

DEFF Research Database (Denmark)

Bjerrum, Peter; Olsen, Kasper Nefer

2001-01-01

Inform@ed space Sensorial Perception And Computer Enchancement - bidrag til Nordisk Arkitekturforskningsforenings IT-konference, AAA april 2001.......Inform@ed space Sensorial Perception And Computer Enchancement - bidrag til Nordisk Arkitekturforskningsforenings IT-konference, AAA april 2001....

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 Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

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

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

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.

Research of management information system of radiation protection for low temperature nuclear heating reactor

International Nuclear Information System (INIS)

Bai Hongtao; Wang Jiaying; Wu Manxue

2001-01-01

Management information system of radiation protection for low temperature reactor uses computer to manage the data of the low temperature nuclear heating reactor radiation monitoring, it saves the data from the front real-time radiation monitoring system, comparing these data with historical data to give the consequence. Also, the system provides some picture in order to show space information at need. The system, based on Microsoft Access 97, consists of nine parts, including radiation dose, environmental data, meteorological data and so on. The system will have value in safely operation of the low temperature nuclear heating reactor

Problems of space-time behaviour of nuclear reactors

International Nuclear Information System (INIS)

Obradovic, D.

1966-01-01

This paper covers a review of literature and mathematical methods applied for space-time behaviour of nuclear reactors. The review of literature is limited to unresolved problems and trends of actual research in the field of reactor physics [sr

Information Space, Information Field, Information Environment

Directory of Open Access Journals (Sweden)

Victor Ya. Tsvetkov

2014-08-01

Full Text Available The article analyzes information space, information field and information environment; shows that information space can be natural and artificial; information field is substantive and processual object and articulates the space property; information environment is concerned with some object and acts as the surrounding in relation to it and is considered with regard to it. It enables to define information environment as a subset of information space. It defines its passive description. Information environment can also be defined as a subset of information field. It corresponds to its active description.

Space reactor preliminary mechanical design

International Nuclear Information System (INIS)

Meier, K.L.

1983-01-01

An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO 2 block core mounts, bolted collar fuel module restraints, and a BeO central plug

Space reactor electric systems: system integration studies, Phase 1 report

International Nuclear Information System (INIS)

Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

1983-01-01

This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

Heat transfer capability analysis of heat pipe for space reactor

International Nuclear Information System (INIS)

Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang

2015-01-01

To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

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

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Enabling autonomous control for space reactor power systems

International Nuclear Information System (INIS)

Wood, R. T.

2006-01-01

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

Radionuclide inventories for short run-time space nuclear reactor systems

International Nuclear Information System (INIS)

Coats, R.L.

1993-01-01

Space Nuclear Reactor Systems, especially those used for propulsion, often have expected operation run times much shorter than those for land-based nuclear power plants. This produces substantially different radionuclide inventories to be considered in the safety analyses of space nuclear systems. This presentation describes an analysis utilizing ORIGEN2 and DKPOWER to provide comparisons among representative land-based and space systems. These comparisons enable early, conceptual considerations of safety issues and features in the preliminary design phases of operational systems, test facilities, and operations by identifying differences between the requirements for space systems and the established practice for land-based power systems. Early indications are that separation distance is much more effective as a safety measure for space nuclear systems than for power reactors because greater decay of the radionuclide activity occurs during the time to transport the inventory a given distance. In addition, the inventories of long-lived actinides are very low for space reactor systems

Nuclear reactor refuelable in space

International Nuclear Information System (INIS)

El-Genk, M.S.; Buden, D.; Mims, J.E.

1992-01-01

This patent describes a gas cooled nuclear reactor suitable for use in space. It comprises a lightweight structure comprising a plurality of at least three sections, each sector comprising a container for a reactor core separate and distinct from the reactor cores of the other sectors, each sector being capable of operating on its own and in cooperation with one or more of the other sectors and each sector having a common juncture with every other structure; and means associated with each sector independently introducing gas coolant into and extracting coolant from each sector to cool the core therein, wherein in event of failure of the cooling system of a core in a sector, one or more of the other sectors comprise means for conducting heat away from the failed sector core and means for convecting the heat away, and wherein operation of the one or more other sectors is maintained

Assessment of nuclear reactor concepts for low power space applications

Science.gov (United States)

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.

White Paper – Use of LEU for a Space Reactor

Energy Technology Data Exchange (ETDEWEB)

Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-11

Historically space reactors flown or designed for the U.S. and Russia used Highly Enriched Uranium (HEU) for fuel. HEU almost always produces a small and lighter reactor. Since mass increases launch costs or decreases science payloads, HEU was the natural choice. However in today’s environment, the proliferation of HEU has become a major concern for the U.S. government and hence a policy issue. In addition, launch costs are being reduced as the space community moves toward commercial launch vehicles. HEU also carries a heavy security cost to process, test, transport and launch. Together these issues have called for a re-investigation into space reactors the use Low Enriched Uranium (LEU) fuel.

Space nuclear reactors: energy gateway into the next millennium

International Nuclear Information System (INIS)

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

1981-01-01

Power - reliable, abundant and economic - is the key to man's conquest of the Solar System. Space activities of the next few decades will be highlighted by the creation of the extraterrestrial phase of human civilization. Nuclear power is needed both to propel massive quantities of materials through cislunar and eventually translunar space, and to power the sophisticated satellites, space platforms, and space stations of tomorrow. To meet these anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100-kW(e) heat pipe nuclear reactor. The objectives of this program are to develop components for a space nuclear power plant capable of unattended operation for 7 to 10 years; having a reliability of greater than 0.95; and weighing less than 1910 kg. In addition, this heat pipe reactor is also compatible for launch by the US Space Transportation System

Space-time neutronic analysis of postulated LOCA's in CANDU reactors

International Nuclear Information System (INIS)

Luxat, J.C.; Frescura, G.M.

1978-01-01

Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

Analytic solutions of the multigroup space-time reactor kinetics equations

International Nuclear Information System (INIS)

Lee, C.E.; Rottler, S.

1986-01-01

The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)

An assessment of space reactor technology needs and recommendations for development

International Nuclear Information System (INIS)

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

1996-01-01

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. sm-bullet Level 0 emdash General Trends and Issues sm-bullet Level 1 emdash Generic Space Capabilities to Address Trends sm-bullet Level 2 emdash Requirements to Support Capabilities sm-bullet Level 3 emdash System Types Capable of Meeting Requirements sm-bullet Level 4 emdash Generic Reactor System Types sm-bullet Level 5 emdash 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. copyright 1996 American Institute of Physics

Proposal of space reactor for nuclear electric propulsion system

International Nuclear Information System (INIS)

Nishiyama, Takaaki; Nagata, Hidetaka; Nakashima, Hideki

2009-01-01

A nuclear reactor installed in spacecrafts is considered here. The nuclear reactor could stably provide an enough amount of electric power in deep space missions. Most of the nuclear reactors that have been developed up to now in the United States and the former Soviet Union have used uranium with 90% enrichment of 235 U as a fuel. On the other hand, in Japan, because the uranium that can be used is enriched to below 20%, the miniaturization of the reactor core is difficult. A Light-water nuclear reactor is an exception that could make the reactor core small. Then, the reactor core composition and characteristic are evaluated for the cases with the enrichment of the uranium fuel as 20%. We take up here Graphite reactor, Light-water reactor, and Sodium-cooled one. (author)

Evaluating Russian space nuclear reactor technology for United States applications

International Nuclear Information System (INIS)

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

1994-01-01

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

Pellet bed reactor for multi-modal space power

International Nuclear Information System (INIS)

Buden, D.; Williams, K.; Mast, P.; Mims, J.

1987-01-01

A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in the tens of megawatts. The electrical power requirements are envisioned to be twofold: long-duration lower power levels will be needed for station keeping, communications, and/or surveillance; short-duration higher power levels will be required for pulsed power devices. These power characteristics led to the proposal of a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multimegawatt reactor power source are standby, alert, and pulsed power modes; high-thermal output heat source (approximately 1000 MWt peak power); long lifetime station keeping power (10 to 30 years); high temperature output (1500 K to 1800 K); rapid-burst power transition; high reliability (above 95 percent); and stringent safety standards compliance. The proposed pellet bed reactor is designed to satisfy these characteristics

Approach to developing reliable space reactor power systems

International Nuclear Information System (INIS)

Mondt, J.F.; Shinbrot, C.H.

1991-01-01

The Space Reactor Power System Project is in the engineering development phase of a three-phase program. During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described in this paper along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top down systems approach which includes a point design based on a detailed technical specification of a 100 kW power system

Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

International Nuclear Information System (INIS)

Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

2016-01-01

Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

National Research Council Canada - National Science Library

Presby, Andrew L

2004-01-01

.... This has potential benefits for space nuclear reactor power systems currently in development. The primary obstacle to space operation of thermophotovoltaic devices appears to be the low heat rejection temperatures which necessitate large radiator areas...

A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

1998-01-01

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

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

Science.gov (United States)

Bowles, K. J.

1974-01-01

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

Autonomous Control of Space Reactor Systems

International Nuclear Information System (INIS)

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

2007-01-01

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 available to perform intelligent control functions that are necessary for both normal and abnormal operational conditions

Autonomous Control of Space Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

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.

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

International Nuclear Information System (INIS)

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

2004-01-01

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 1.4x107 n/s at the core-reflector interface, with an incoming current of 3.4x106 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 a

Cermet-fueled reactors for advanced space applications

International Nuclear Information System (INIS)

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

Reactor Operations informal monthly report December 1994

International Nuclear Information System (INIS)

1994-12-01

Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

Space-reactor electric systems: subsystem technology assessment

International Nuclear Information System (INIS)

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

1983-01-01

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

DESIGN SAFETY PROBLEMS OF NUCLEAR REACTORS IN SPACE FOR ELECTRICAL POWER

Energy Technology Data Exchange (ETDEWEB)

Pickler, D A

1963-06-15

A general treatment is presented of some of the problems in the design safety of reactors which are to be operated in space. The basic requirements of these reachigh temperatures. The usual concept of a space reactor is described briefly, and the hazards of an assumed unmanned vehicle with an enriched-U-fueled reactor are examined during its launching, orbit, and reentry. Graphs are given for the dose vs distance downwind for an excursion of 100 Mw-sec, for the activity vs time after shutdown of a reactor which has been operated for 5 yr at 100 kw(t), and for the altitude vs orbital lifetime. Apparent conflicts between the basic requirements are discussed. (D.L.C.)

Primary loop simulation of the SP-100 space nuclear reactor

International Nuclear Information System (INIS)

Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.

2011-01-01

Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

Autonomous Control Capabilities for Space Reactor Power Systems

International Nuclear Information System (INIS)

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

2004-01-01

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

Shielding considerations for advanced space nuclear reactor systems

International Nuclear Information System (INIS)

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 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

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

Science.gov (United States)

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.

Molten-salt reactor information system

International Nuclear Information System (INIS)

Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

1975-06-01

The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

The IAEA power reactor information system - PRIS

International Nuclear Information System (INIS)

Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

1983-01-01

The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

Improved Nuclear Reactor and Shield Mass Model for Space Applications

Science.gov (United States)

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.

Applying design principles to fusion reactor configurations for propulsion in space

International Nuclear Information System (INIS)

Carpenter, S.A.; Deveny, M.E.; Schulze, N.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. Three design principles (DP's) were applied 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. A preliminary rating of these configurations was performed, and it was concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS)

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

International Nuclear Information System (INIS)

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

1987-01-01

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets 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

Solid-Core, Gas-Cooled Reactor for Space and Surface Power

International Nuclear Information System (INIS)

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

2006-01-01

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Status report on nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-01-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

INFORMATION SPACE– EDUCATIONAL SPACE

Directory of Open Access Journals (Sweden)

Monica LIA

2015-11-01

Full Text Available This paper has set the objective of researching how education is influenced by the information society. The first step was to define more precisely the information space. The second step was to identify how information space intersects with the family space and institutional space educational levels represented by pre-school / school and pre-university (kindergarten, at elementary / middle school / high school. Interrelationship between the above mentioned areas was another objective of the research. All these elements have been investigated through the original intention to identify how the information space can become an educational tool to support the family space, education and institutional space. Also, the aim of this research is to offer some solutions in this regard. Often the educational efforts appear to be blocked by the existence of this space. But this paper demonstrates that Informational space can be an enemy of the educational system or can support systems if we knew the internal structure and mechanisms. We can make the Informational Space to work in order to accomplish the educational scope.

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

1996-01-01

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

1989-12-01

This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

Nuclear reactor descriptions for space power systems analysis

Science.gov (United States)

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

1972-01-01

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

Space nuclear reactor system diagnosis: Knowledge-based approach

International Nuclear Information System (INIS)

Ting, Y.T.D.

1990-01-01

SP-100 space nuclear reactor system development is a joint effort by the Department of Energy, the Department of Defense and the National Aeronautics and Space Administration. The system is designed to operate in isolation for many years, and is possibly subject to little or no remote maintenance. This dissertation proposes a knowledge based diagnostic system which, in principle, can diagnose the faults which can either cause reactor shutdown or lead to another serious problem. This framework in general can be applied to the fully specified system if detailed design information becomes available. The set of faults considered herein is identified based on heuristic knowledge about the system operation. The suitable approach to diagnostic problem solving is proposed after investigating the most prevalent methodologies in Artificial Intelligence as well as the causal analysis of the system. Deep causal knowledge modeling based on digraph, fault-tree or logic flowgraph methodology would present a need for some knowledge representation to handle the time dependent system behavior. A proposed qualitative temporal knowledge modeling methodology, using rules with specified time delay among the process variables, has been proposed and is used to develop the diagnostic sufficient rule set. The rule set has been modified by using a time zone approach to have a robust system design. The sufficient rule set is transformed to a sufficient and necessary one by searching the whole knowledge base. Qualitative data analysis is proposed in analyzing the measured data if in a real time situation. An expert system shell - Intelligence Compiler is used to develop the prototype system. Frames are used for the process variables. Forward chaining rules are used in monitoring and backward chaining rules are used in diagnosis

Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

Spiegelberg, R.

1992-01-01

The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

A preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems

International Nuclear Information System (INIS)

Parlos, A.G.; Khan, E.U.; Frymire, R.; Negron, S.; Thomas, J.K.; Peddicord, K.L.

1991-01-01

Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the devleopment of highly compact space reactors. The calculations reported in this study include a neutronic design trade-off study using a two-dimensioinal neutron transport model, as well as a simplified one-dimensional thermal analysis of the reactor core. In arriving at the most desirable configuration, various options have been considered and analyzed, and their advantages/disadvantages have been compared. However, because of space limitation, only the most favorable reactor configuration is presented in this summary

Cermet-fueled reactors for multimegawatt space power applications

International Nuclear Information System (INIS)

Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

1988-01-01

The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Lewis, B.R.; Klein, A.C.; Pawlowski, R.A.

1993-01-01

Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range

Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F.

2015-01-01

The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW e and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k eff = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

2015-07-01

The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

Power reactor information system (PRIS)

International Nuclear Information System (INIS)

1989-06-01

Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

Power conditioning for space nuclear reactor systems

Science.gov (United States)

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.

Integrated design support systems for conceptual design of a space power reactor

International Nuclear Information System (INIS)

Shimoda, Hiroshi; Yoshikawa, Hidekazu; Takahashi, Makoto; Takeoka, Satoshi; Nagamatsu, Takashi; Ishizaki, Hiroaki

1999-01-01

In the process of conceptual design of large and complex engineering systems such as a nuclear power reactor, there must be various human works by several fields of engineers on each stage of design, analysis and evaluation. In this study, we have rearranged the design information to reduce the human workloads and have studied an efficient method to support the conceptual design works by new information technologies. For this purpose, we have developed two design support environments for conceptual design of a space power reactor as a concrete design target. When constructing an integrated design support environment, VINDS, which employs virtual reality(VR) technology, we focused on visualization of physical structure, functional organization and analysis calculation with full usage of easy perception and direct manipulation of VR. On the other hand, when constructing another asynchronous and distributed design support environment, WINDS, which employs WWW technology, we improved the support functions for cooperative design works among various fields of experts. In this paper, the basic concepts, configurations and functions of the design support environments are first described, then the future improvement is surveyed by their intercomparison. (author)

Performance analysis of Brayton cycle system for space power reactor

International Nuclear Information System (INIS)

Li Zhi; Yang Xiaoyong; Zhao Gang; Wang Jie; Zhang Zuoyi

2017-01-01

The closed Brayton cycle system now is the potential choice as the power conversion system for High Temperature Gas-cooled Reactors because of its high energy conversion efficiency and compact configuration. The helium is the best working fluid for the system for its chemical stability and small neutron absorption cross section. However, the Helium has small mole mass and big specific volume, which would lead to larger pipes and heat exchanger. What's more, the big compressor enthalpy rise of helium would also lead to an unacceptably large number of compressor's stage. For space use, it's more important to satisfy the limit of the system's volume and mass, instead of the requirement of the system's thermal capacity. So Noble-Gas binary mixture of helium and xenon is presented as the working fluid for space Brayton cycle. This paper makes a mathematical model for space Brayton cycle system by Fortran language, then analyzes the binary mixture of helium and xenon's properties and effects on power conversion units of the space power reactor, which would be helpful to understand and design the space power reactor. The results show that xenon would lead to a worse system's thermodynamic property, the cycle's efficiency and specific power decrease as xenon's mole fraction increasing. On the other hand, proper amount of xenon would decrease the enthalpy changes in turbomachines, which would be good for turbomachines' design. Another optimization method – the specific power optimization is also proposed to make a comparison. (author)

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

International Nuclear Information System (INIS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-01-01

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

Gaseous-fuel nuclear reactor research for multimegawatt power in space

Science.gov (United States)

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.

Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems

International Nuclear Information System (INIS)

1994-12-01

The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors

Information Architecture the Design of Digital Information Spaces

CERN Document Server

Ding, Wei

2009-01-01

Information Architecture is about organizing and simplifying information, designing and integrating information spaces/systems, and creating ways for people to find and interact with information content. Its goal is to help people understand and manage information and make right decisions accordingly. In the ever-changing social, organizational and technological contexts, Information Architects not only design individual information spaces (e.g., individual websites, software applications, and mobile devices), but also tackle strategic aggregation and integration of multiple information spaces

Safety program considerations for space nuclear reactor systems

International Nuclear Information System (INIS)

Cropp, L.O.

1984-08-01

This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given

Russian RBMK reactor design information

International Nuclear Information System (INIS)

1993-11-01

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received

Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

International Nuclear Information System (INIS)

Koenig, D.R.; Gido, R.G.; Brandon, D.I.

1985-01-01

The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

Multidimensional space-time kinetics of a heavy water moderated nuclear reactor

International Nuclear Information System (INIS)

Winn, W.G.; Baumann, N.P.; Jewell, C.E.

1980-01-01

Diffusion theory analysis of a series of multidimensional space-time experiments is appraised in terms of the final experiment of the series. In particular, TRIMHX diffusion calculations were examined for an experiment involving free-fall insertion of a 235 U-bearing rod into a heavy water moderated reactor with a large reflector. The experimental transient flux-tilts were accurately reproduced after cross section adjustments forced agreement between static diffusion calculations and static reactor measurements. The time-dependent features were particularly well modeled, and the bulk of the small discrepancies in space-dependent features should be removable by more refined cross-section adjustments. This experiment concludes a series of space-time experiments that span a wide range of delayed neutron holdback effects. TRIMHX calculations of these experiments demonstrate the accuracy of the modeling employed in the code

Neutronics Study of the KANUTER Space Propulsion Reactor

International Nuclear Information System (INIS)

Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee

2014-01-01

The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe 135 , and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity

Neutronics Study of the KANUTER Space Propulsion Reactor

Energy Technology Data Exchange (ETDEWEB)

Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe{sup 135}, and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity.

Environmental Information Document: L-reactor reactivation

International Nuclear Information System (INIS)

Mackey, H.E. Jr.

1982-04-01

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program

Environmental Information Document: L-reactor reactivation

Energy Technology Data Exchange (ETDEWEB)

Mackey, H.E. Jr. (comp.)

1982-04-01

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

Drug Information in Space Medicine

Science.gov (United States)

Bayuse, Tina M.

2009-01-01

Published drug information is widely available for terrestrial conditions. However, information on dosing, administration, drug interactions, stability, and side effects is scant as it relates to use in Space Medicine. Multinational crews on board the International Space Station present additional challenges for drug information because medication nomenclature, information available for the drug as well as the intended use for the drug is not standard across countries. This presentation will look at unique needs for drug information and how the information is managed in Space Medicine. A review was conducted of the drug information requests submitted to the Johnson Space Center Pharmacy by Space Medicine practitioners, astronaut crewmembers and researchers. The information requested was defined and cataloged. A list of references used was maintained. The wide range of information was identified. Due to the information needs for the medications in the on-board medical kits, the Drug Monograph Project was created. A standard method for answering specific drug information questions was generated and maintained by the Johnson Space Center Pharmacy. The Drug Monograph Project will be presented. Topic-centered requests, including multinational drug information, drug-induced adverse reactions, and medication events due to the environment will be highlighted. Information management of the drug information will be explained. Future considerations for drug information needs will be outlined.

Space reactor/organic Rankine conversion - A near-term state-of-the-art solution

Science.gov (United States)

Niggemann, R. E.; Lacey, D.

The use of demonstrated reactor technology with organic Rankine cycle (ORC) power conversion can provide a low cost, minimal risk approach to reactor-powered electrical generation systems in the near term. Several reactor technologies, including zirconium hydride, EBR-II and LMFBR, have demonstrated long life and suitability for space application at the operating temperature required by an efficient ORC engine. While this approach would not replace the high temperature space reactor systems presently under development, it could be available in a nearer time frame at a low and predictable cost, allowing some missions requiring high power levels to be flown prior to the availability of advanced systems with lower specific mass. Although this system has relatively high efficiency, the heat rejection temperature is low, requiring a large radiator on the order of 3.4 sq m/kWe. Therefore, a deployable heat pipe radiator configuration will be required.

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

International Nuclear Information System (INIS)

Berg, Thomas A.; Disney, Richard K.

2004-01-01

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

Development of a Robust Tri-Carbide Fueled Reactor for Multi-Megawatt Space Power and Propulsion Applications

International Nuclear Information System (INIS)

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

2004-01-01

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

Spacing grid intended for nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Patterson, J.F.; Flora, B.S.

1977-01-01

This invention concerns a new improved type of spacing grid that can be used in nuclear reactor fuel assemblies. Under the invention a spacing grid is provided, preferably of the bimetallic type. This grid includes a set of flexible inconel strips positioned by structural 'zircalloy' fittings, having relatively low neutron absorption characteristics in comparison with systems where the flexible strips are welded in position, or where the spring forms an integral part of the structure. The openings for the fuel elements which are defined by the structural fittings intercrossing are fitted internally with bosses which work in conjunction with a spring directed downwards as from the flexible strip so as to position the individual fuel rods in their respective openings inside the grid structure. These flexible strips are arranged in rows extending in directions which depend on the particular design of the fuel asembly and which contain flexible components so distributed that the loads of the individual springs tend to equalize each other mutually. The reaction load exerting itself on the supporting structure is reduced to the minimum, and this results in a lesser distortion in the reactor and an equalisation of the spring loads [fr

Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

Energy Technology Data Exchange (ETDEWEB)

Washington, K.E.

1986-05-01

The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations.

Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

International Nuclear Information System (INIS)

Washington, K.E.

1986-05-01

The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1982-06-01

This publication is the third in a proposed series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. Subsequent issues in this series will update the information contained herein. This publication contains basically three kinds of information: (1) routes approved by the Commission for the shipment of irradiated reactor fuel, (2) information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and (3) cumulative amounts of material shipped

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

Science.gov (United States)

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.

Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

International Nuclear Information System (INIS)

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

2004-01-01

Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

Study of space reactors for exploration missions

Energy Technology Data Exchange (ETDEWEB)

Cliquet, Elisa; Ruault, Jean-Marc; Masson, Frederic, E-mail: elisa.cliquet@cnes.fr, E-mail: frederic.masson@cnes.fr [Centre National d' Etudes Spatiales (CNES), Paris (France); Roux, Jean-Pierre; Paris, Nicolas; Cazale, Brice; Manifacier, Laurent, E-mail: jean-pierre.roux@areva.com [AREVA TA, Aix en Provence, (France); Poinot-Salanon, Christine, E-mail: christine.poinot@cea.fr [Comissariado a l' Energie Atomique et Aux Energies alternatives (CEA), Paris (France)

2013-07-01

Nuclear propulsion has been studied for many decades. The power density of nuclear fission is much higher than chemical process, and for missions to outer solar system requiring several hundred of kilowatts, or for flexible manned missions to Mars requiring several megawatts, nuclear electric propulsion might be the only option offering a reasonable mass in low earth orbit. Despite the existence of low power experiences - SNAP10 in the 60's or Buk/Topaz in the 60-80's - no high power reactor has been developed: investment cost, long term time frame, high technological challenges and radioactive hazards are the main challenges we must overtake. However, it seems reasonable to look at the technical challenges that have to be overcome for a next generation of nuclear electric systems for space exploration. This paper will present some recent studies going on in France, on space reactors for exploration. Three classes of power have been considered: 10kWe, 100kWe, and several megawatts. Available data from previous studies and developments performed in Russia, USA], and Europe, have been collected and gave us a large overview of potential technical solutions. This was the starting point of a trade-off analysis aiming at the selection of the best options, with regards to the technological readiness level in France and Europe. The resulting preliminary designs will be presented and critical technologies needing maturation activities will be highlighted. (author)

Study of space reactors for exploration missions

International Nuclear Information System (INIS)

Cliquet, Elisa; Ruault, Jean-Marc; Masson, Frederic; Roux, Jean-Pierre; Paris, Nicolas; Cazale, Brice; Manifacier, Laurent; Poinot-Salanon, Christine

2013-01-01

Nuclear propulsion has been studied for many decades. The power density of nuclear fission is much higher than chemical process, and for missions to outer solar system requiring several hundred of kilowatts, or for flexible manned missions to Mars requiring several megawatts, nuclear electric propulsion might be the only option offering a reasonable mass in low earth orbit. Despite the existence of low power experiences - SNAP10 in the 60's or Buk/Topaz in the 60-80's - no high power reactor has been developed: investment cost, long term time frame, high technological challenges and radioactive hazards are the main challenges we must overtake. However, it seems reasonable to look at the technical challenges that have to be overcome for a next generation of nuclear electric systems for space exploration. This paper will present some recent studies going on in France, on space reactors for exploration. Three classes of power have been considered: 10kWe, 100kWe, and several megawatts. Available data from previous studies and developments performed in Russia, USA], and Europe, have been collected and gave us a large overview of potential technical solutions. This was the starting point of a trade-off analysis aiming at the selection of the best options, with regards to the technological readiness level in France and Europe. The resulting preliminary designs will be presented and critical technologies needing maturation activities will be highlighted. (author)

Performance testing of refractory alloy-clad fuel elements for space reactors

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

1985-01-01

Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

Small space reactor power systems for unmanned solar system exploration missions

International Nuclear Information System (INIS)

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

Space dependence of reactivity parameters on reactor dynamic perturbation measurements

International Nuclear Information System (INIS)

Maletti, R.; Ziegenbein, D.

1985-01-01

Practical application of reactor-dynamic perturbation measurements for on-power determination of differential reactivity weight of control rods and power coefficients of reactivity has shown a significant dependence of parameters on the position of outcore detectors. The space dependence of neutron flux signal in the core of a VVER-440-type reactor was measured by means of 60 self-powered neutron detectors. The greatest neutron flux alterations are located close to moved control rods and in height of the perturbation position. By means of computations, detector positions can be found in the core in which the one-point model is almost valid. (author)

Neutronics characteristics of space power reactors

International Nuclear Information System (INIS)

Little, W.; Barner, J.

1986-01-01

The objective of the paper is to describe the neutronic characteristics of a range of possible space reactor designs, and indicate the relative advantages and disadvantages of the various designs. Fuel designs to be considered are cermets (i.e., ceramic particles embedded in a metal matrix) consisting of UO 2 or Nn ceramic particles in matrices of Nb, Mo, Ta, or W. These cermet fuels are compared to a UN pin-type design. UN was selected for the reference fuel material since it has a somewhat higher density than UO 2 (i.e., 14.32 versus 10.96 gm/cc), which allows a lower minimum critical mass for both ceramic and cermet designs

Effects of irradiation on properties of refractory alloys with emphasis on space power reactor applications

International Nuclear Information System (INIS)

Wiffen, F.W.

1984-01-01

The probable effects of irradiation on niobium and tungsten alloys in use as components of thermionic convertors in a space reactor were reviewed by the author in 1971. While considerably more data on refractory metals have been generated since that time, the data have not been reviewed with respect to space reactor applications. This paper attempts such a review. The approach used is to work from the most recently available review of irradiation effects for each alloy system (where such a review is available) and to discuss that review and more recent data judged to be the most useful in establishing likely behavior in high-temperature reactor service. 28 figures, 6 tables

Selection of power plant elements for future reactor space electric power systems

International Nuclear Information System (INIS)

Buden, D.; Bennett, G.A.; Copper, K.

1979-09-01

Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

Systems aspects of a space nuclear reactor power system

Science.gov (United States)

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.

Space nuclear reactor concepts for avoidance of a single point failure

International Nuclear Information System (INIS)

El-Genk, M. S.

2007-01-01

This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

McKissock, B.I.; Bloomfield, H.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. The 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. (author)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

Mckissock, B.I.; Bloomfield, H.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

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

International Nuclear Information System (INIS)

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

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

Science.gov (United States)

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.

Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

International Nuclear Information System (INIS)

Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

1987-01-01

Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

International Nuclear Information System (INIS)

Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

1964-01-01

1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

The museum as information space

DEFF Research Database (Denmark)

Navarrete, T.; Mackenzie Owen, J.

2016-01-01

space to being outside the museum in the online information space of the Internet. This has fundamental implications for the institutional role of museums, our understanding of metadata and the methods of documentation. The onsite museum institution will, eventually, not be able to function...... as an institutional entity on the Internet, for in this new information space, objects, collections and museums, all function as independent components in a vast universe of data, side by side at everyone’s disposal at anytime. Potentially, users can access cultural heritage anytime, anywhere and anyhow. © The Author......Although museums vary in nature and may have been founded for all sorts of reasons, central to all museum institutions are the collected objects. These objects are information carriers organized in a catalogue system. In this chapter, the museum will be conceived as an information space, consisting...

14. informal meeting on reactor noise

International Nuclear Information System (INIS)

1981-01-01

The present booklet contains abstracts of papers from the 14th informal meeting on reactor noise held at St. Englmar in April 1981. The main topics dealt with are vibration and loose part monitoring, leak detection, noise theory and noise applications and in the final part data processing and pattern recognition techniques. (orig.)

The 'MELUSINE' reactor at Grenoble, France, and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the MELUSINE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities and specialized irradiation devices (loops and capsules). The information is presented in the form of six information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities

SP-100 space reactor power system readiness

International Nuclear Information System (INIS)

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

1992-01-01

This paper discusses the SP-100 Space Reactor Power System which 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). 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

State-space representation of the reactor dynamics equations

International Nuclear Information System (INIS)

Bernard, J.A.

1995-01-01

This paper describes a novel formulation of the reactor space-independent kinetics equations. The intent is to present these equations in a form that is both compatible with modern control theory and mathematically rigorous. It is desired to write the kinetics equations in the standard state variable representation, x = Ax, where x is the state vector and A is the system matrix and, at the same time, avoid mathematical compromises such as the linearization of an equation about a particular operating point. The advantage to this proposed formulation is that it may allow the lateral transfer of existing control concepts, some that have been developed for other fields, to the operation of nuclear reactors. For example, sliding mode control has been developed to allow robots to function in a robust manner in the presence of changes in the system model. This is necessary because a robot is expected to be capable of picking up an object of unknown mass and moving that object along a specified trajectory. The variability of the object's mass introduces an uncertainty into the system model that is used to deduce the appropriate control action. Thus, the robot controller must be made robust against such variations. Sliding mode control is one means of accomplishing this. A reactor controller might benefit from the same concept if its objective were to cause the reactor power to move along a demanded trajectory despite the presence of some uncertainty in the net amount of reactivity that is present

Recommendations for space reactor R ampersand D tasks

International Nuclear Information System (INIS)

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

1995-01-01

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

Mechanical characterization of superalloys for space reactors

International Nuclear Information System (INIS)

Duchesne, J.

1989-01-01

The aim of this work is the selection of structural materials that can be used in the temperature range 600-900 0 C for a gas cooled space reactor producing electricity. Superalloys fit best the temperature range required. Five nickel base alloys are chosen for their good mechanical behaviour: HAYNES 230, HASTELLOY S, HASTELLOY X, HASTELLOY XR and PYRAD 38D. Metallography, tensile and hardness tests are realized. Sample contraction is evidenced for some creep tests, under low stress: 20MPa at 800 0 C, on HAYNES 230 and HASTELLOY X, probably related to the structural evolution of these materials corresponding to a decrease of the crystal parameter [fr

Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

International Nuclear Information System (INIS)

Fletcher, C.D.

1986-01-01

The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M., E-mail: guimarae@ieav.cta.br, E-mail: lamartine.guimaraes@pq.cnpq.br, E-mail: jamil@ieav.cta.br, E-mail: jalnsgf@outlook.com, E-mail: borges.em@hotmail.com, E-mail: ecorborges@hotmail.com, E-mail: ivayolini@gmail.com, E-mail: guilherme_placco@ig.com.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil); Barrios Junior, Ary Garcia, E-mail: arygarcia89@yahoo.com [Faculdade de Tecnologia Sao Francisco (FATESF), Jacarei, SP (Brazil)

2013-07-01

This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

International Nuclear Information System (INIS)

Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M.; Barrios Junior, Ary Garcia

2013-01-01

This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

Advanced micro-reactor for space and deep sea exploration: a scientific Brazilian vision

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca; Lobo, Paulo D.C.

2011-01-01

Humankind is at the point to initiate a new adventure in its evolutionary journey, the colonization of other planets of our solar system and space travels. Also, there is still another frontier where the human presence is scarce, the oceans and the Earth seabed. To have success in the exploration of these new frontiers a fundamental requirement must be satisfied: secure availability of energy for life support and others processes. This work deals with the establishment of a basis for a Brazilian nuclear research and development (R and D) program to develop micro-reactor (MR) technologies that may be used in the seabed, the space or another hostile environment on Earth. The work presents a set of basic requirements that is used to define the best reactor type to be used in these environments. Also, the limits and dimensions that define the class of micro-reactors are discussed. The fast neutron spectrum was chosen as the best for the MR and the limits for the active core volume and thermal power are 30 liters and 5 MW. (author)

Space reactors. Progress report, October 1981-March 1982

International Nuclear Information System (INIS)

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 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

Research on Monte Carlo improved quasi-static method for reactor space-time dynamics

International Nuclear Information System (INIS)

Xu Qi; Wang Kan; Li Shirui; Yu Ganglin

2013-01-01

With large time steps, improved quasi-static (IQS) method can improve the calculation speed for reactor dynamic simulations. The Monte Carlo IQS method was proposed in this paper, combining the advantages of both the IQS method and MC method. Thus, the Monte Carlo IQS method is beneficial for solving space-time dynamics problems of new concept reactors. Based on the theory of IQS, Monte Carlo algorithms for calculating adjoint neutron flux, reactor kinetic parameters and shape function were designed and realized. A simple Monte Carlo IQS code and a corresponding diffusion IQS code were developed, which were used for verification of the Monte Carlo IQS method. (authors)

State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

Energy Technology Data Exchange (ETDEWEB)

Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

2017-02-15

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Spacing grids for a fuel pencil bundle in a nuclear reactor assembly

International Nuclear Information System (INIS)

Feutrel, Claude.

1977-01-01

This invention relates to the lattices forming the spacing of a bundle of clad fuel pencils in a nuclear reactor assembly, particularly in a water cooled or fast reactor, the purpose of such lattices being to maintain these pencils parallel with respect to each other and according to a given lattice arrangement, whilst also providing these pencils with a flexible support according to different successive areas apportioned with their length in order to present them from vibrating under the effect of the circulation of a liquid coolant environment flowing in contact with these pencils [fr

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

Energy Technology Data Exchange (ETDEWEB)

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.

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

International Nuclear Information System (INIS)

Buden, D.

1993-01-01

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

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

Helmick, H.H.; Schwenk, F.C.

1978-01-01

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

Risk-informed design of a pebble bed gas reactor

International Nuclear Information System (INIS)

Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

2003-01-01

One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

Public information circular for shipments of irradiated reactor fuel. Revision 4

International Nuclear Information System (INIS)

1984-06-01

This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1983-07-01

This publication contains basically three kinds of information: routes approved by the Commission for the shipment of irradiated reactor fuel, information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and cumulative amounts of material shipped

The 'SILOE' reactor at Grenoble, France and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the SILOE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

The 'OSIRIS' reactor at Saclay, France and available hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the OSIRIS reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Design of a management information system for the Shielding Experimental Reactor ageing management

International Nuclear Information System (INIS)

He Jie; Xu Xianhong

2010-01-01

The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

Design of a management information system for the Shielding Experimental Reactor ageing management

Energy Technology Data Exchange (ETDEWEB)

He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

2010-01-15

The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

Brayton rotating units for space reactor power systems

Energy Technology Data Exchange (ETDEWEB)

Gallo, Bruno M.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies and Chemical and Nuclear Engineering Dept., The Univ. of New Mexico, Albuquerque, NM 87131 (United States)

2009-09-15

Designs and analyses models of centrifugal-flow compressor and radial-inflow turbine of 40.8kW{sub e} Brayton Rotating Units (BRUs) are developed for 15 and 40 g/mole He-Xe working fluids. Also presented are the performance results of a space power system with segmented, gas cooled fission reactor heat source and three Closed Brayton Cycle loops, each with a separate BRU. The calculated performance parameters of the BRUs and the reactor power system are for shaft rotational speed of 30-55 krpm, reactor thermal power of 120-471kW{sub th}, and turbine inlet temperature of 900-1149 K. With 40 g/mole He-Xe, a power system peak thermal efficiency of 26% is achieved at rotation speed of 45 krpm, compressor and turbine inlet temperatures of 400 and 1149 K and 0.93 MPa at exit of the compressor. The corresponding system electric power is 122.4kW{sub e}, working fluid flow rate is 1.85 kg/s and the pressure ratio and polytropic efficiency are 1.5% and 86.3% for the compressor and 1.42% and 94.1% for the turbine. For the same nominal electrical power of 122.4kW{sub e}, decreasing the molecular weight of the working fluid (15 g/mole) decreases its flow rate to 1.03 kg/s and increases the system pressure to 1.2 MPa. (author)

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

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

Science.gov (United States)

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.

Graphic-object information system {open_quotes}research base for reactor materials science{close_quotes}

Energy Technology Data Exchange (ETDEWEB)

Markina, N.V.; Lebedeva, E.E.; Arkhangel`skii, N.V.; Semenov, S.B.; Moiseev, A.L.

1994-11-01

An information system developed for reactor materials research is described. The information system incorporates an expert system, MATREKS, and a heirarchial data base. The data base contains information from 20 Russian research reactors. The information system structure, data base structure, search methods, system output modes, and technical facilities and software required are briefly discussed. 6 refs., 2 figs.

Preliminary neutronic design of spock reactor: A nuclear system for space power generation

International Nuclear Information System (INIS)

Burgio, N.; Santagata, A.; Cumo, M.; Fasano, A.; Frullini, M.

2007-01-01

Aim of this paper is to preliminary investigates the neutronic features of an upgrade of the MAUS [1] nuclear reactor whose core will be able to supply a thermoelectric converter in order to generate 30 kW of electricity for space applications. The neutronic layout of SPOCK (Space Power Core Ka) is a compact, MOX fuelled, liquid metal cooled and totally reflected fast reactor with a control system based on neutron absorption. Spock, that during the heart and launch operation must be maintained in sub-critical state, has to start up in the outer space at 40 K temperatures with the coolant in a solid state and it will reach the operating steady condition at the maximum temperature of 1300 K with the coolant in the liquid state. The main design goal is to maintains, in the operating conditions of a typical space mission, the control of the appropriate criticality margin versus temperature and coolant physical state. For this purpose, a neutronic/thermal-hydraulic calculation chain able to assists the entire design process must be set up. As preliminary recognition, MCNPX 2.5.0 and FLUENT calculations were carried out. The emerging key features of SPOCK are: an equilateral triangular mesh of 91 cylindrical UO 2 fuel rods with a Molybdenum clad ensured by two grids of the same material, cooled by liquid Sodium and contained in an AISI 316 L vessel. The core is totally wrapped by a Beryllium reflector that hosts six absorber (B 4 C) rotating control rods. The reactor shape is cylindrical (radius = 30 cm and height = 60 cm) with a total mass of 275 kg. The excess reactivity was of 5000 PCM at 1300 K. A preliminary evaluation of the control rods worth and a power spatial distribution were also discussed. Through the definition of an ideal reference K e ff value at 300 K for the actual SPOCK configuration, a sensitivity analysis on various cross sections data and material physical properties was performed for the given mission temperature range, allowing consideration on

Problems of space-time behaviour of nuclear reactors; Problemi prostorno-vremenskog ponasanja nuklearnih reaktora

Energy Technology Data Exchange (ETDEWEB)

Obradovic, D [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

1966-07-01

This paper covers a review of literature and mathematical methods applied for space-time behaviour of nuclear reactors. The review of literature is limited to unresolved problems and trends of actual research in the field of reactor physics. Dat je pregled literature i matematickih metoda koje se koriste prilikom tretiranja prostorno-vremenskog ponasanja nuklearnih reaktora. Pregled literature ogranicen je na jos neresene probleme i pravce u kojima su danas usmerena istrazivanja u ovoj oblasti fizike nuklearnih reaktora (author)

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

International Nuclear Information System (INIS)

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

The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Anderson, Robert Stephen [Idaho National Laboratory

2015-12-01

Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologies may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.

Risk-informed design guidance for future reactor systems

International Nuclear Information System (INIS)

Delaney, Michael J.; Apostolakis, George E.; Driscoll, Michael J.

2005-01-01

Future reactor designs face an uncertain regulatory environment. It is anticipated that there will be some level of probabilistic insights in the regulations and supporting regulatory documents for Generation-IV nuclear reactors. Central to current regulations are design basis accidents (DBAs) and the general design criteria (GDC), which were established before probabilistic risk assessments (PRAs) were developed. These regulations implement a structuralist approach to safety through traditional defense in depth and large safety margins. In a rationalist approach to safety, accident frequencies are quantified and protective measures are introduced to make these frequencies acceptably low. Both approaches have advantages and disadvantages and future reactor design and licensing processes will have to implement a hybrid approach. This paper presents an iterative four-step risk-informed methodology to guide the design of future-reactor systems using a gas-cooled fast reactor emergency core cooling system as an example. This methodology helps designers to analyze alternative designs under potential risk-informed regulations and to anticipate design justifications the regulator may require during the licensing process. The analysis demonstrated the importance of common-cause failures and the need for guidance on how to change the quantitative impact of these potential failures on the frequency of accident sequences as the design changes. Deliberation is an important part of the four-step methodology because it supplements the quantitative results by allowing the inclusion in the design choice of elements such as best design practices and ease of online maintenance, which usually cannot be quantified. The case study showed that, in some instances, the structuralist and the rationalist approaches were inconsistent. In particular, GDC 35 treats the double-ended break of the largest pipe in the reactor coolant system with concurrent loss of offsite power and a single

The DIDO-reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the DIDO reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

TUTORIAL SUPPORT IN THE INFORMATION SPACE OF SCHOOLS

Directory of Open Access Journals (Sweden)

Вита Иммануиловна Глизбург

2017-12-01

Full Text Available The article deals with teachers and professional competence tyutorskoy, tutor support schools in the information space. The main requirements to the tutor support in the information space of schools, conditions of readiness of the teacher to tutor support schools in the information space. A theoretical analysis of the concept of information competence, tyutorskoy competence. It is noted that in thestructure of information competence must exist an element associated with the motivation, need and interest in the acquisition of knowledge and skills in the field of technical, software and information. Formulated key performance indicators definition of information competence of the tutor. The authors noted that information and work with it is in the modern educational and information space school mandatory components of pedagogical activity as a system.Analyzed and presented on the basis of the author’s experience of the possibility of using information and educational Moodle shell with tutor support positions in the information space of schools, sets out the basic elements and resources Moodle shell with a description of the implementation. In particular, the disclosed educational resources information and educational shell Moodle for remote support learning. The article describes a model developed by the author’s tutor support schools in the information space.

Design and evaluation of materials for space reactors

International Nuclear Information System (INIS)

Tavassoli, A.A.; Vrillon, B.; Robert, G.

1990-01-01

The French programme envisages a 20 kWe reactor, project ERATO, with three technological options. The first option is a sodium cooled reactor, derived from the fast breeder reactor technology, (upper core outlet temperature of 700 0 C). The second option is based on the High Temperature Gas-cooled Reactor technology (outlet temperature range 700 0 C-900 0 C). The third option is the reference solution, lithium cooled and UN fuelled fast spectrum reactor, (outlet temperature as high as 1200 0 C). The choice is essentially dominated by material considerations, and more specifically by the problems related to the compatibility with the cooling medium and to the high temperature creep resistance. For the first system limited work will be needed as the technology used is well experimented and there is a wealth of information on the austenitic stainless steel Type 316L-SPH. For the second system, most of the work has been concentrated on characterization of existing commercial alloys. This has included the preselection and the testing of a number of superalloys irradiated or not. The results obtained from high temperature tensile and creep tests have allowed selection of Haynes 230 as the primary candidate material and have also permitted calculation of allowable design stresses for this alloy. For the very high temperature system the French R and D programme has focused on Mo-Re alloys. The results obtained to this date from microstructural examinations and mechanical tests performed on different alloy compositions have allowed selection of Mo-25%Re for future optimization work. They have also shown the need for evaluation of creep properties at low stresses where microstructural instabilities are likely to occur as a result of long exposure to high temperature

Thermofluid-neutronic stability of the rotating, fluidized bed, space-power reactor

International Nuclear Information System (INIS)

Lee, C.C.; Jones, O.C.; Becker, M.

1993-01-01

A rotating fluidized bed nuclear reactor has the potential of being a vary attractive option for ultra-high power space systems, especially for propulsion. Research has already examined fuel bed expansion due to variations in state variables, propellant flow rate, and rotational speed, and has also considered problems related to thermal stress. This paper describes the results of a coupled thermofluid-neutronic analysis where perturbations in fuel bed height caused by maneuvering changes in operating conditions alter power levels due to varying absorption of neutrons which would otherwise leak from the system, mainly through the nozzle. This first analysis was not a detailed stability analysis. Rather, it utilized simplified neutronic methods, and was intended to provide an order-of-magnitude assessment of the stability of the reactor with the intention to determine whether or not stability might be a 'concept killer'. Stability was compared with a fixed-fuel-bed reactor of identical geometry for three different cases comprising a set of small, medium and large sizes/powers from 250 MW to 5 GW. It was found that power fluctuations in the fluidized bed reactor were larger by 100 db or more than expected in a packed bed reactor of the same geometry, but never resulted in power excursions. Margins to unit gain in some cases, however, were sufficiently small that the approximations in this quasi-2-dimensional model may not be sufficiently accurate to preclude significant excursions. (orig.)

Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; van Dam, H.; Kleiss, E.B.J.; van Uitert, G.C.; Veldhuis, D.

1982-01-01

The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations.

Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1982-01-01

The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations. (author)

PRIS-STATISTICS: Power Reactor Information System Statistical Reports. User's Manual

International Nuclear Information System (INIS)

2013-01-01

The IAEA developed the Power Reactor Information System (PRIS)-Statistics application to assist PRIS end users with generating statistical reports from PRIS data. Statistical reports provide an overview of the status, specification and performance results of every nuclear power reactor in the world. This user's manual was prepared to facilitate the use of the PRIS-Statistics application and to provide guidelines and detailed information for each report in the application. Statistical reports support analyses of nuclear power development and strategies, and the evaluation of nuclear power plant performance. The PRIS database can be used for comprehensive trend analyses and benchmarking against best performers and industrial standards.

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Designing informal learning spaces using student perspectives

Directory of Open Access Journals (Sweden)

Matthew David Riddle

2012-06-01

Full Text Available This article describes the design of informal learning spaces at an Australian university that support students in the generation of knowledge. Recent learning space design projects at La Trobe have been informed by a number of pre-existing projects, including a small research project on student use of technologies, a national project on learning space design, and a significant curriculum renewal process at the university. It demonstrates the ways in which evidence based on student perspectives and principles developed through applied research in teaching and learning can inform real world learning space design projects in a higher education context.

Transactions of the nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1991-10-01

This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Design of particle bed reactors for the space nuclear thermal propulsion program

International Nuclear Information System (INIS)

Ludewig, H.; Powell, J.R.; Todosow, M.; Maise, G.; Barletta, R.; Schweitzer, D.G.

1996-01-01

This paper describes the design for the Particle Bed Reactor (PBR) that was considered for the Space Nuclear Thermal Propulsion (SNTP) Program. The methods of analysis and their validation are outlined first. Monte Carlo methods were used for the physics analysis, several new algorithms were developed for the fluid dynamics, heat transfer and transient analysis; and commercial codes were used for the stress analysis. We carried out a critical experiment, prototypic of the PBR to validate the reactor physics; blowdown experiments with beds of prototypic dimensions were undertaken to validate the power-extraction capabilities from particle beds. In addition, materials and mechanical design concepts for the fuel elements were experimentally validated. (author)

Transactions of the Twentieth Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-10-01

This report contains summaries of papers on reactor safety research to be presented at the 20th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 21--23, 1992. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included

Transactions of the twenty-fifth water reactor safety information meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

Transactions of the twenty-fifth water reactor safety information meeting

International Nuclear Information System (INIS)

Monteleone, S.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

The Vehicular Information Space Framework

Science.gov (United States)

Prinz, Vivian; Schlichter, Johann; Schweiger, Benno

Vehicular networks are distributed, self-organizing and highly mobile ad hoc networks. They allow for providing drivers with up-to-the-minute information about their environment. Therefore, they are expected to be a decisive future enabler for enhancing driving comfort and safety. This article introduces the Vehicular Information Space framework (VIS). Vehicles running the VIS form a kind of distributed database. It enables them to provide information like existing hazards, parking spaces or traffic densities in a location aware and fully distributed manner. In addition, vehicles can retrieve, modify and delete these information items. The underlying algorithm is based on features derived from existing structured Peer-to-Peer algorithms and extended to suit the specific characteristics of highly mobile ad hoc networks. We present, implement and simulate the VIS using a motorway and an urban traffic environment. Simulation studies on VIS message occurrence show that the VIS implies reasonable traffic overhead. Also, overall VIS message traffic is independent from the number of information items provided.

Incomplete information and fractal phase space

International Nuclear Information System (INIS)

Wang, Qiuping A.

2004-01-01

The incomplete statistics for complex systems is characterized by a so called incompleteness parameter ω which equals unity when information is completely accessible to our treatment. This paper is devoted to the discussion of the incompleteness of accessible information and of the physical signification of ω on the basis of fractal phase space. ω is shown to be proportional to the fractal dimension of the phase space and can be linked to the phase volume expansion and information growth during the scale refining process

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

International Nuclear Information System (INIS)

Samim Anghaie

2002-01-01

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

The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the DR 2 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of seven information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

SHOVAV-JUEL. A one dimensional space-time kinetic code for pebble-bed high-temperature reactors with temperature and Xenon feedback

International Nuclear Information System (INIS)

Nabbi, R.; Meister, G.; Finken, R.; Haben, M.

1982-09-01

The present report describes the modelling basis and the structure of the neutron kinetics-code SHOVAV-Juel. Information for users is given regarding the application of the code and the generation of the input data. SHOVAV-Juel is a one-dimensional space-time-code based on a multigroup diffusion approach for four energy groups and six groups of delayed neutrons. It has been developed for the analysis of the transient behaviour of high temperature reactors with pebble-bed core. The reactor core is modelled by horizontal segments to which different materials compositions can be assigned. The temperature dependence of the reactivity is taken into account by using temperature dependent neutron cross sections. For the simulation of transients in an extended time range the time dependence of the reactivity absorption by Xenon-135 is taken into account. (orig./RW)

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

Science.gov (United States)

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.

Analysis of space-time core dynamics on reactor accident at Chernobyl

International Nuclear Information System (INIS)

Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

1987-05-01

Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

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.

Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

International Nuclear Information System (INIS)

Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

1974-01-01

In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

Transactions of the twenty-second water reactor safety information meeting

International Nuclear Information System (INIS)

1994-10-01

This report contains summaries of papers on reactor safety research to be presented at the 22nd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 24--26, 1994. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session. Individual papers have been cataloged separately

PC version of PRIS (Power Reactor Information System)

International Nuclear Information System (INIS)

Fukala, J.; Stanik, Z.; White, D.

1990-05-01

The IAEA has been collecting operating experience data on nuclear power plants in the Member States since 1970. In 1980 a computerized database was established, the IAEA Power Reactor Information System (PRIS). To make PRIS data available to the Member States in a more convenient format, the development of a PC version of PRIS started in 1989

FBR and RBR particle bed space reactors

International Nuclear Information System (INIS)

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 0 K), high coolant-outlet temperatures (1500 to 3000 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 2 -cooled mode. The RBR will operate only in the open-cycle H 2 -cooled mode

Reactor power peaking information display

International Nuclear Information System (INIS)

Book, T.L.; Kochendarfer, R.A.

1986-01-01

This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

Designing informal learning spaces using student perspectives

OpenAIRE

Matthew David Riddle; Kay Souter

2012-01-01

This article describes the design of informal learning spaces at an Australian university that support students in the generation of knowledge. Recent learning space design projects at La Trobe have been informed by a number of pre-existing projects, including a small research project on student use of technologies, a national project on learning space design, and a significant curriculum renewal process at the university. It demonstrates the ways in which evidence based on student perspectiv...

The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Earth and space science information systems

Energy Technology Data Exchange (ETDEWEB)

Zygielbaum, A. (ed.) (Jet Propulsion Laboratory, California Institute of Technology, Pasadena, California 91109 (United States))

1993-01-01

These proceedings represent papers presented at the Earth and Space Science Information Systems (ESSIS) Conference. The attendees included scientists and engineers across many disciplines. New trends in information organizations were reviewed. One hundred and twenty eight papers are included in this volume, out of these two have been abstracted for the Energy Science and Technology database. The topics covered in the papers range from Earth science and technology to astronomy and space, planetary science and education. (AIP)

Risk-informing special treatment requirements for reactors

International Nuclear Information System (INIS)

McKenna, E.M.; Reed, T.A.

2001-01-01

The U.S. Nuclear Regulatory Commission (NRC) is proposing to make regulatory changes to the scope of structures, systems, and components (SSCs) requiring special treatment. ''Special treatment requirements'' refers to those specific examples of regulations that are applied in order to provide a high degree of assurance that SSC will be capable of performing their intended functions when needed. The current scope of SSCs covered by the special treatment requirements governing commercial nuclear reactors is deterministically based and stems primarily from the evaluation of selected design basis events, as described in updated final safety analysis reports (UFSARs). This regulatory framework provides reasonable assurance of no undue risk to the health and safety of the public. However, recent advances in technology, coupled with operating reactor experience, have suggested that an alternative approach that would use a risk-informed process for evaluating SSC safety significance, would, in turn, result in a more focused determination of which SSCs should receive special treatment requirements. (author)

Two-phase reduced gravity experiments for a space reactor design

International Nuclear Information System (INIS)

Antoniak, Z.I.

1986-08-01

Future space missions envision the use of large nuclear reactors utilizing either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed, to coordinate all ongoing and planned reduced gravity flow experiments

Loss of coolant accident mitigation for liquid metal cooled space reactors

International Nuclear Information System (INIS)

Georgevich, Vladimir; Best, Frederick; Erdman, Carl

1989-01-01

A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

Fossil fuel furnace reactor

Science.gov (United States)

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.

Determination of space-energy distribution of resonance neutrons in reactor lattice cell and calculation of resonance integrals

International Nuclear Information System (INIS)

Zmijarevic, I.

1980-01-01

Space-energy distribution of resonance neutrons in reactor lattice cell was determined by solving the Boltzmann equation by spherical harmonics method applying P-3 approximation. Computer code SPLET used for these calculations is described. Resonance absorption and calculation of resonance integrals are described as well. Effective resonance integral values for U-238 resonance at 6.7 Ev are calculated for heavy water reactor cell with metal, oxide and carbide fuel elements

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

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

Demonstration-informative center based on research reactor IR-50 in heat regime

International Nuclear Information System (INIS)

Krupenina, Ph.

2000-01-01

Many problems exist in the nuclear field, but the most significant one is the public's mistrust of Nuclear Energy. Strong downfalls of the radiological culture affect public perception, the main paradox being the situation after Chernobyl. The task of creating a Demonstration-Informative Center (Minatom RF) on reactor IR-50 research is conducted by Research and Development Institute of Power Engineering (ENTEK). The IR-50 is situated on the grounds of the institute. It will be a unique event when the functional reactor is situated in the center of the city. The purposes of the Demonstration-Informative Center are discussed. (authors)

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat 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; (g) a bed of insulating material 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; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

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

Science.gov (United States)

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.

Nuclear power in space. Use of reactors and radioactive substances as power sources in satellites and space probes

International Nuclear Information System (INIS)

Hoestbaeck, Lars

2008-11-01

Today solar panels are the most common technique to supply power to satellites. Solar panels will work as long as the power demand of the satellite is limited and the satellite can be equipped with enough panels, and kept in an orbit that allows enough sunlight to hit the panels. There are various types of space missions that do not fulfil these criteria. With nuclear power these types of missions can be powered regardless of the sunlight and as early as 1961 the first satellite with a nuclear power source was placed in orbit. Out of seventy known space missions that has made use of nuclear power, ten have had some kind of failure. In no case has the failure been associated with the nuclear technology used. This report discusses to what degree satellites with nuclear power are a source for potential radioactive contamination of Swedish territory. It is not a discussion for or against nuclear power in space. Neither is it an assessment of consequences if radioactive material from a satellite would reach the earth's surface. Historically two different kinds of Nuclear Power Sources (NPS) have been used to generate electric power in space. The first is the reactor where the energy is derived from nuclear fission of 235 U and the second is the Radioisotope Thermoelectric Generator (RTG) where electricity is generated from the heat of naturally decaying radionuclides. NPS has historically only been used in space by United States and the Soviet Union (and in one failing operation Russia). Nuclear Power Sources have been used in three types of space objects: satellites, space probes and moon/Mars vehicles. USA has launched one experimental reactor into orbit, all other use of NPS by the USA has been RTG:s. The Soviet Union, in contrast, only launched a few RTG:s but nearly forty reactors. The Soviet use of NPS is less transparent than the use in USA and some data published on Soviet systems are more or less well substantiated assessments. It is likely that also future

Reinforced confinement in a nuclear reactor

International Nuclear Information System (INIS)

Norman, H.

1988-01-01

The present invention concerns a nuclear reactor containing a reactor core, a swimming pool space that is filled and pressurized with a neutron-absorbing solution, a reactor tank, at least one heat exchanger, at least one inlet line, at least one return line and at least one circulation pump, where the said reactor tank is confined in the said swimming pool space and designed to be cooled with the aid of relatively pure water, which is fed by means of the said at least one circulating pump to the said reactor tank from the said heat exchanger via the said at least one inlet line and is returned to the heat exchanger via the said at least one return line. The problem that is to be solved by the invention is to design a reactor of the above type in such a way that a complete confinement of the primary circuit of the reactor is achieved at relatively low extra cost. This problem is solved by providing the reactor with a special confinement space that confines the heat exchanger, but not the reactor tank, with the confinement space and the swimming pool space being fashioned in the same concrete body

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

International Nuclear Information System (INIS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

Flow of ideal fluid through a central region of a nuclear reactor wire-spaced fuel subassembly

International Nuclear Information System (INIS)

Schmid, J.

1991-04-01

The results are given of calculations of the flow of an ideal fluid through the central region of a nuclear reactor wire-spaced fuel subassembly. The computer code used is briefly described. (author). 10 figs., 4 refs

Space Propulsion via Spherical Torus Fusion Reactor

International Nuclear Information System (INIS)

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

2003-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 204 days, with an initial mass in low Earth orbit of 1630 mt. Engineering conceptual design, analysis, and assessment were performed on all major systems including nuclear fusion reactor, magnetic nozzle, power conversion, fast wave plasma heating, fuel pellet injector, startup/re-start fission reactor and battery, and other systems. Detailed fusion reactor design included analysis of plasma characteristics, power balance and utilization, first wall, toroidal field coils, heat transfer, and neutron/X-ray radiation

Plant with nuclear reactor, in particular a thermal reactor

International Nuclear Information System (INIS)

Straub, H.

1988-01-01

The reactor core of the plant has tubular and vertically movable control rods moved by a flow of coolant under pressure. Each control rod surrounds a similarly tubular guide rod, stationary relative to the reactor core, leaving an annular slot-like space therebetween. The inside of each guide rod forms a first pressure chamber supplied with the coolant under pressure. The upper end of each control rod is closed and has a vertical shaft that extends into the inside of the guide rod and forms therewith a second annular slot-like space. At least one first restriction is provided in the first annular slot-like space and at least one second restriction is provided in the second annular slot-like space. A second pressure chamber is formed between both restrictions. The coolant supplied to the guide rod thus returns to the pressure vessel surrounding the reactor core through the second annular slot-like space, the second pressure chamber and the first annular slot-like space. Controlling means are provided, with which pressure thrusts can be generated if necessary in the coolant within the first pressure chamber. (author) 5 refs., 10 figs

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

Science.gov (United States)

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

1990-01-01

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

Concepts for space nuclear multi-mode reactors

International Nuclear Information System (INIS)

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

1983-01-01

A number of nuclear multi-mode reactor power plants are conceptualized for use with solid core, fixed particle bed and rotating particle bed reactors. Multi-mode systems generate high peak electrical power in the open cycle mode, with MHD generator or turbogenerator converters and cryogenically stored coolants. Low level stationkeeping power and auxiliary reactor cooling (i.e., for the removal of reactor afterheat) are provided in a closed cycle mode. Depending on reactor design, heat transfer to the low power converters can be accomplished by heat pipes, liquid metal coolants or high pressure gas coolants. Candidate low power conversion cycles include Brayton turbogenerator, Rankine turbogenerator, thermoelectric and thermionic approaches. A methodology is suggested for estimating the system mass of multi-mode nuclear power plants as a function of peak electric power level and required mission run time. The masses of closed cycle nuclear and open cycle chemical power systems are briefly examined to identify the regime of superiority for nuclear multi-mode systems. Key research and technology issues for such power plants are also identified

Towards Mobile Information Systems for Indoor Space

Directory of Open Access Journals (Sweden)

Xiaoxiang Zhang

2016-01-01

Full Text Available With the rapid development of Internet of things (IOT and indoor positioning technologies such as Wi-Fi and RFID, indoor mobile information systems have become a new research hotspot. Based on the unique features of indoor space and urgent needs on indoor mobile applications, in this paper we analyze some key issues in indoor mobile information systems, including positioning technologies in indoor environments, representation models for indoor spaces, query processing techniques for indoor moving objects, and index structures for indoor mobile applications. Then, we present an indoor mobile information management system named IndoorDB. Finally, we give some future research topics about indoor mobile information systems.

The FRJ 1 reactor (MERLIN) at Juelich, F.R. Germany and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the FRJ 1 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

Preliminary closed Brayton cycle study for a space reactor application

International Nuclear Information System (INIS)

Guimaraes, Lamartine Nogueira Frutuoso; Carvalho, Ricardo Pinto de; Camillo, Giannino Ponchio

2007-01-01

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 to establish a starting concept for the CBCL components specifications, and to develop a demonstrative simulator of CBCL in nominal operation conditions. The ENU/IEAv 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, CO2 and gas mixtures such as helium and xenon. At this moment the simulator is running with Helium as the working fluid. Simplified models of heat and mass transfer are being developed to simulate thermal components. Future efforts will focus on keeping track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. (author)

Preliminary closed Brayton cycle study for a space reactor application

Energy Technology Data Exchange (ETDEWEB)

Guimaraes, Lamartine Nogueira Frutuoso; Carvalho, Ricardo Pinto de [Institute for Advanced Studies, Sao Jose dos Campos, SP (Brazil)]. E-mail: guimarae@ieav.cta.br; Camillo, Giannino Ponchio [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil)]. E-mail: gianninocamillo@gmail.com

2007-07-01

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 to establish a starting concept for the CBCL components specifications, and to develop a demonstrative simulator of CBCL in nominal operation conditions. The ENU/IEAv 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, CO2 and gas mixtures such as helium and xenon. At this moment the simulator is running with Helium as the working fluid. Simplified models of heat and mass transfer are being developed to simulate thermal components. Future efforts will focus on keeping track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. (author)

Intelligent information database of the thermal-hydraulic characteristics for a future marine water reactor

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki

2000-01-01

At the Ship Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated type marine water reactor has been conducted. This current study aims at developing an intelligent information database program with the thermal-hydraulic characteristics of a future marine water reactor on the basis of the valuably experimental knowledge, which was obtained from the above-mentioned studies. In this paper, the experimental knowledge with the flow boiling of a once-through steam generator and the natural circulation of primary water under a ship rolling motion was converted into an intelligent information database program. The program was created as a Windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability for any helical-coil type once-through steam generator design, (2) reference and graphic display of the experimental data, (3) reference of the information such as analysis method and experimental apparatus. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized reactor with helical-coil type steam generator. (author)

Design Options to Reduce Development Cost of First Generation Surface Reactors

International Nuclear Information System (INIS)

Poston, David I.; Marcille, Thomas F.

2006-01-01

Low-power surface reactors have the potential to have the lowest development cost of any space reactor application, primarily because system alpha (mass/kg) is not of utmost importance and mission lifetimes do not have to be a decade or more. Even then, the development cost of a surface reactor can vary substantially depending on the performance requirements (e.g. mass, power, lifetime, reliability) and technical development risk deemed acceptable by the end-user. It is important for potential users to be aware of these relationships before they determine their future architecture (i.e. decide what they need). Generally, the greatest potential costs of a space reactor program are a nuclear-powered ground test and extensive material development campaigns, so it is important to consider options that can minimize the need for or complexity of such tasks. The intended goal of this paper is to inform potential surface reactor users of the potential sensitivities of surface reactor development cost to design requirements, and areas where technical risk can be traded with development cost

Development Program of the Advanced HANARO Reactor in Korea

International Nuclear Information System (INIS)

Yang, I.-S.; Ahn, J.-H.; Han, K.-I.; Parh, C.; Jun, B.-J.; Kim, Y.-J.

2006-01-01

The development program of an advanced HANARO (AHR) reactor started in Korea to keep abreast of the increasing future demand, from both home and abroad, for research activities. This paper provides a review of the status of research reactors in Korea, the operating experience of the HANARO, the design principles and preliminary features of an advanced HANARO reactor, and the specific strategy of an advanced HANARO reactor development program. The design principles were established in order to design a new multi-purpose research reactor that is safe, economically competitive and technically feasible. These include the adaptation of the HANARO design concept, its operating experience, a high ratio of flux to power, a high degree of safety, improved economic efficiency, improved operability and maintainability, increased space and expandability, and ALARA design optimization. The strategy of an advanced HANARO reactor development program considers items such as providing a digital advanced HANARO reactor in cyber space, a method for the improving the design quality and economy of research reactors by using Computer Integrated Engineering, and more effective advertising using diverse virtual reality. This development program will be useful for promoting the understanding of and interest in the operating HANARO as well as an advanced HANARO reactor under development in Korea. It will provide very useful information to a country that may need a research reactor in the near future for the promotion of public health, bio-technology, drug design, pharmacology, material processing, and the development of new materials. (author)

Heat pipe nuclear reactor for space power

Science.gov (United States)

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.

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

Science.gov (United States)

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

2004-02-01

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

EMERIS: an advanced information system for a materials testing reactor

International Nuclear Information System (INIS)

Adorjan, F.; Buerger, L.; Lux, I.; Mesko, L.; Szabo, K.; Vegh, J.; Ivanov, V.V.; Mozhaev, A.A.; Yakovlev, V.V.

1990-06-01

The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

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

Science.gov (United States)

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.

Adaptability of Brayton cycle conversion systems to fast, epithermal and thermal spectrum space nuclear reactors

International Nuclear Information System (INIS)

Tilliette, Z.P.

1988-01-01

The two French Government Agencies C.N.E.S. (Centre National d'Etudes Spatiales) and C.E.A. (Commissariat a l'Energie Atomique) are carrying out joint preliminary studies on space nuclear power systems for future ARIANE 5 launch vehicle applications. The Brayton cycle is the reference conversion system, whether the heat source is a liquid metal-cooled (NaK, Na or Li) reactor or a gas-cooled direct cycle concept. The search for an adequate utilization of this energy conversion means has prompted additional evaluations featuring the definition of satisfactory cycle conditions for these various kinds of reactor concepts. In addition to firstly studied fast and epithermal spectrum ones, thermal spectrum reactors can offer an opportunity of bringing out some distinctive features of the Brayton cycle, in particular for the temperature conditioning of the efficient metal hydrides (ZrH, Li/sub 7/H) moderators. One of the purposes of the paper is to confirm the potential of long lifetime ZrH moderated reactors associated with a gas cycle and to assess the thermodynamical consequences for both Nak(Na)-cooled or gas-cooled nuclear heat sources. This investigation is complemented by the definition of appropriate reactor arrangements which could be presented on a further occasion

Thermo-mechanical behaviour of FBTR reactor vessel due to natural convection in cover gas space

International Nuclear Information System (INIS)

Srinivasan, G.; Varadarajan, S.; Kapoor, R.P.

1988-01-01

Fast Breeder Test Reactor is a 40 MW(t), loop type sodium cooled reactor, similar in design to Rapsodie. The Reactor Assembly, which is the heart of FBTR, comprises the Reactor Vessel (RV) housed in a safety vessel within a concrete cell (A1 Cell). The RV which supports the core is shielded at the top by two rotatable plugs which are stacked with layers of borated graphite and steel. The smaller plug (SRP), is mounted excentric to the larger one (LRP). A nominal annular gap of 16 mm is provided between RV and LRP and between LRP and SRP to enable free rotation of the plugs. Stainless Steel insulation is fixed inside the steel vessel, to avoid overheating of the A1 Cell concrete. The core is supported by the Grid Plate (GP), bolted to the RV. During preheating, sodium charging and isothermal runs upto 350 0 C, temperature asymmetries were noticed in the reactor vessel wall in the cover gas space. This was attributable to convection currents in the annulus between RV and LRP. The asymmetries also resulted in a lateral shift of the grid plate. This paper discusses our experience in suppressing these convection currents, and minimising the grid plate shift

Space reactors - What is a kilogram

International Nuclear Information System (INIS)

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

1984-01-01

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

Space reactors: What is a kilogram

International Nuclear Information System (INIS)

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

Advanced nuclear reactor types and technologies

International Nuclear Information System (INIS)

Ignatiev, V.; Devell, L.

1995-01-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

Advanced nuclear reactor types and technologies

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1995-07-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

Space reactor fuel element testing in upgraded TREAT

International Nuclear Information System (INIS)

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

1993-01-01

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

Space reactor fuel element testing in upgraded TREAT

Science.gov (United States)

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

1993-01-01

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

SPLET - A program for calculating the space-lethargy distribution of epithermal neutrons in a reactor lattice cell

International Nuclear Information System (INIS)

Matausek, M.V.; Zmijatevic, I.

1981-01-01

A procedure to solve the space-single-lethargy dependent transport equation for epithermal neutrons in a cylindricised multi-region reactor lattice cell has been developed and proposed in the earlier papers. Here, the computational algorithm is comprised and the computing program SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, as well as the related integral quantities as reaction rates and resonance integrals, is described. (author)

The TERRA project, a space nuclear micro-reactor case study

International Nuclear Information System (INIS)

Guimaraes, Lamartine N.F.; Nascimento, Jamil A.; Borges, Eduardo M.; Lobo, Paulo D. Castro; Placco, Guilherme M.; Barrios Junior, Ary G.

2011-01-01

The TEcnologia de Reatores Rapidos Avancados project, also known as TERRA Project is been conducted by the Institute for Advanced Studies IEAv. The TERRA project has a general objective of understanding and developing the key technologies that will allow (Brazil) the use of nuclear technology to generate electricity in space. This electricity may power several space systems and/or a type of plasma based engine. Also, the type of reactor intended for space may be used for power generation in very inhospitable environment such as the ocean floor. Some of the mentioned technologies may include: Brayton cycles, Stirling engines, heat pipes and its coupled systems, nuclear fuel technology, new materials and several others. Once there is no mission into which apply this technology, at this moment, this research may be conducted in many forms and ways. The fact remains that when this technology becomes needed there will be no way that we (Brazilians) will be able to buy it from. This technology, in this sense, is highly strategic and will be the key to commercially explore deep space. Therefore, there is the need to face the development problems and solve them, to gain experience with our own rights and wrongs. This paper will give a brief overview of what has been done so far, on experimental facilities and hardware that could support space system development, including a Brayton cycle test facility, Tesla turbine testing, and Stirling engine development and modeling. Our great problem today is lack of human resources. To attend that problem we are starting a new graduate program that will allow overcoming that, given the proper time frame. (author)

The TERRA project, a space nuclear micro-reactor case study

Energy Technology Data Exchange (ETDEWEB)

Guimaraes, Lamartine N.F.; Nascimento, Jamil A.; Borges, Eduardo M.; Lobo, Paulo D. Castro, E-mail: guimarae@ieav.cta.br, E-mail: jamil@ieav.cta.br, E-mail: eduardo@ieav.cta.br [Divisao de Energia Nuclear. Instituto de Estudos Avancados, Sao Jose dos Campos, SP (Brazil); Placco, Guilherme M.; Barrios Junior, Ary G. [Faculdade de Tecnologia Sao Francisco (FATESF), Jacarei, SP (Brazil)

2011-07-01

The TEcnologia de Reatores Rapidos Avancados project, also known as TERRA Project is been conducted by the Institute for Advanced Studies IEAv. The TERRA project has a general objective of understanding and developing the key technologies that will allow (Brazil) the use of nuclear technology to generate electricity in space. This electricity may power several space systems and/or a type of plasma based engine. Also, the type of reactor intended for space may be used for power generation in very inhospitable environment such as the ocean floor. Some of the mentioned technologies may include: Brayton cycles, Stirling engines, heat pipes and its coupled systems, nuclear fuel technology, new materials and several others. Once there is no mission into which apply this technology, at this moment, this research may be conducted in many forms and ways. The fact remains that when this technology becomes needed there will be no way that we (Brazilians) will be able to buy it from. This technology, in this sense, is highly strategic and will be the key to commercially explore deep space. Therefore, there is the need to face the development problems and solve them, to gain experience with our own rights and wrongs. This paper will give a brief overview of what has been done so far, on experimental facilities and hardware that could support space system development, including a Brayton cycle test facility, Tesla turbine testing, and Stirling engine development and modeling. Our great problem today is lack of human resources. To attend that problem we are starting a new graduate program that will allow overcoming that, given the proper time frame. (author)

Return and profitability of space programs. Information - the main product of flights in space

Science.gov (United States)

Nikolova, Irena

The basic branch providing global information, as a product on the market, is astronautics and in particular aero and space flights. Nowadays economic categories like profitability, return, and self-financing are added to space information. The activity in the space information service market niche is an opportunity for realization of high economic efficiency and profitability. The present report aims at examining the possibilities for return and profitability of space programs. Specialists in economics from different countries strive for defining the economic effect of implementing space technologies in the technical branches on earth. Still the priorities here belong to government and insufficient market organization and orientation is apparent. Attracting private investors and searching for new mechanisms of financing are the factors for increasing economic efficiency and return of capital invested in the mentioned sphere. Return of utilized means is an economically justified goal, a motive for a bigger enlargement of efforts and directions for implementing the achievements of astronautics in the branches of economy on earth.

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Analysis and evaluation of ZPPR critical experiments for a 100 kilowatt-electric space reactor

International Nuclear Information System (INIS)

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.

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 needed for fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further reductions in the mass of the SP-100. 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

Status of CEA reactor studies for a 200 kWe turboelectric Space Power System

International Nuclear Information System (INIS)

Carre, F.; Gervaise, F.; Proust, E.; Schwartz, J.P.; Tilliette, Z.; Vrillon, B.

1986-01-01

A reference design for a 200 kWe Space Nuclear Power System has been developed by the CNES and CEA Agencies of the French Government in order to assess within a first study phase running from mid 1984 to mid 1986, the key feasibility issues and the development cost of a Space Power System compatible with the version of the European launcher (ARIANE V), that will be available after 1995, and with adequate power range and lifetime performances for the missions considered at that time. The heat from a fast spectrum lithium cooled reactor is converted by a turboelectric system, selected for its technological readiness and for its advantage over thermionics and thermoelectricity, of minimizing the total mass of 100 to 300 kWe power systems, considering the available radiator area afforded by the specific ARIANE V geometrical features. A heat pipe radiator is preferred to an equivalent gas cooled system, for the increased reliability brought by the large number of independent cooling elements. The successive topics addressed in the paper, include a description of the system main components and steady state operating conditions, and the present views about the start up procedure and the reactor control

Space Power Program Semiannual Progress Report for period ending June 30, 1963

Energy Technology Data Exchange (ETDEWEB)

Miller, A. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

1963-10-11

This is a report of progress on the Oak Ridge National Laboratory's research and development program on nuclear power plants for electrical power production in space vehicles. The work is carried out under AEG Reactor Experiments, Fuels, and Materials, and Reactor Component programs. Research and development work is under way on the stainless steel boiling-potassium reactor and the Medium Power Reactor Experiment, boiling alkali metal heat transfer, high-temperature and refractory alloys, fuel material, and space reactor shielding, particularly in connection with SNAP 2, 8, 10, and 50. Many of these OREL efforts are directed toward the development of a specific type of power plant, but they also furnish a significant contribution of scientific and engineering information needed in other programs on advanced SNAP systems. Progress on research and development directly related to the Medium Power Reactor Experiment (MPRE) is presented mostly in Part I of this report. Progress on the MPRE will, in the future, be reported on a quarterly basis. The form of the reporting will alternate from MPRE Quarterly Progress Reports to Space Power Semiannual Progress Reports.

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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 related 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

NASA space geodesy program: Catalogue of site information

Science.gov (United States)

Bryant, M. A.; Noll, C. E.

1993-01-01

This is the first edition of the NASA Space Geodesy Program: Catalogue of Site Information. This catalogue supersedes all previous versions of the Crustal Dynamics Project: Catalogue of Site Information, last published in May 1989. This document is prepared under the direction of the Space Geodesy and Altimetry Projects Office (SGAPO), Code 920.1, Goddard Space Flight Center. SGAPO has assumed the responsibilities of the Crustal Dynamics Project, which officially ended December 31, 1991. The catalog contains information on all NASA supported sites as well as sites from cooperating international partners. This catalog is designed to provde descriptions and occupation histories of high-accuracy geodetic measuring sites employing space-related techniques. The emphasis of the catalog has been in the past, and continues to be with this edition, station information for facilities and remote locations utilizing the Satellite Laser Ranging (SLR), Lunar Laser Ranging (LLR), and Very Long Baseline Interferometry (VLBI) techniques. With the proliferation of high-quality Global Positioning System (GPS) receivers and Doppler Orbitography and Radiopositioning Integrated by Satellite (DORIS) transponders, many co-located at established SLR and VLBI observatories, the requirement for accurate station and localized survey information for an ever broadening base of scientists and engineers has been recognized. It is our objective to provide accurate station information to scientific groups interested in these facilities.

Meeting the reactor operator's information needs using functional analysis

International Nuclear Information System (INIS)

Nelson, W.R.; Clark, M.T.

1980-01-01

Since the accident at Three Mile Island, many ideas have been proposed for assisting the reactor operator during emergency situations. However, some of the suggested remedies do not alleviate an important shortcoming of the TMI control room: the operators were not presented with the information they needed in a manner which would allow prompt diagnosis of the problem. To address this problem, functional analysis is being applied at the LOFT facility to ensure that the operator's information needs are being met in his procedures and graphic displays. This paper summarizes the current applications of functional analysis at LOFT

Exercise in completing design information questionnaire for model research reactor: model description, notes, questionnaire

International Nuclear Information System (INIS)

Bellinger, J.; Ho, T.

1989-01-01

The document which defines the inspection measures which the IAEA can deploy at any given nuclear facility is known as the Facility Attachment. For the Agency to negotiate an effective Facility Attachment it must have available certain design information, including the facility's identity, capacity and location; the form, location and flow of nuclear material and the layout of important items of equipment; and a description of the features and procedures relating to nuclear material accountancy, containment and surveillance. In practice such information is solicited in a format, standardized for each facility type, known as the Design Information Questionnaire or the D.I.Q. The nuclear activities used as a model in this course are those of a fictitious country called Pacifica. These nuclear activities bear some resemblance to those at the Australian Atomic Energy Commission's Research Establishment at Lucas Heights. Specifically, Pacifica has a 10 MW heavy water cooled and moderated research reactor using enriched uranium fuel which is very similar to the HIFAR reactor. The reactor and the associated laboratories are described and the Design Information Questionnaire for them is completed. figs., tabs

Determining space-energy distribution of thermal neutrons in heterogeneous cylindrically symmetric reactor cell, Master Thesis

International Nuclear Information System (INIS)

Matausek, M. V.

1966-06-01

A combination of multigroup method and P 3 approximation of spherical harmonics method was chosen for calculating space-energy distribution of thermal neutron flux in elementary reactor cell. Application of these methods reduced solution of complicated transport equation to the problem of solving an inhomogeneous system of six ordinary firs-order differential equations. A procedure is proposed which avoids numerical solution and enables analytical solution when applying certain approximations. Based on this approach, computer codes were written for ZUSE-Z-23 computer: SIGMA code for calculating group constants for a given material; MULTI code which uses results of SIGMA code as input and calculates spatial ana energy distribution of thermal neutron flux in a reactor cell. Calculations of thermal neutron spectra for a number of reactor cells were compared to results available from literature. Agreement was satisfactory in all the cases, which proved the correctness of the applied method. Some possibilities for improving the precision and acceleration of the calculation process were found during calculation. (author)

Information space a framework for learning in organizations, institutions and culture

CERN Document Server

Boisot, Max H

2016-01-01

In this book the author lays the foundations for a new political economy of information. The information space, or I-Space is the conceptual framework in which organizations, institutions and cultures are being transformed by new information and communication technologies. In the penultimate chapter, the I-Space's usefulness as an explanatory framework is illustrated with an application: a case study of China's modernization. Information Space proposes a radical shift in the way that we approach the emerging information age and the implications it holds for societies, organizations and individuals.

Systems aspects of a space nuclear reactor power system

International Nuclear Information System (INIS)

Jaffe, L.; Fujita, T.; Beatty, R.

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, attitude 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

Informal Governance in Urban Spaces

DEFF Research Database (Denmark)

Polese, Abel; Rekhviashvili, Lela; Morris, Jeremy Bryan

2017-01-01

Drawing on evidence from the competition for public spaces between street vendors and the authorities in Georgia our contribution through this article is two-fold. First, we provide empirical evidence showing the diverse role of informality in a series of settings, and its capacity to influence d...

Reactor-core-reactivity control device

International Nuclear Information System (INIS)

Miura, Teruo; Sakuranaga, Tomonobu.

1983-01-01

Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

REACTOR SHIELD

Science.gov (United States)

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.

Space and time optimization of nuclear reactors by means of the Pontryagin principle

International Nuclear Information System (INIS)

Anton, V.

1979-01-01

A numerical method is being presented for solving space dependent optimization problems concerning a functional for one dimensional geometries in the few group diffusion approximation. General dimensional analysis was applied to derive relations for the maximum of a functional and the limiting values of the constraints. Two procedures were given for calculating the anisotropic diffusion coefficients in order to improve the results of the diffusion approximation. In this work two procedures were presented for collapsing the microscopic multigroup cross sections, one general and another specific to the space dependent optimization problems solved by means of the Pontryagin maximum principle. Neutron spectrum optimization is performed to ensure the burnup of Pu 239 isotope produced in a thermal nuclear reactor. A procedure is also given for the minimization of finite functional set by means of the Pontryagin maximum principle. A method for determining the characteristics of fission Pseudo products is formulated in one group and multigroup cases. This method is applied in the optimization of the burnup in nuclear reactors with fuel electric cells. A procedure to mjnimze the number of the fuel burnup equations is described. The optimization problems presented and solved in this work point to the efficiency of the maximum principle. Each problem on method presented in the various chapters is accompanied by considerations concerning dual problems and possibilities of further research development. (author)

Fission fragment assisted reactor concept for space propulsion: Foil reactor

International Nuclear Information System (INIS)

Wright, S.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

Temperature and Doppler coefficients of various space nuclear reactors

International Nuclear Information System (INIS)

Mughabghab, S.F.; Ludewig, H. Schmidt, E.

1993-01-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system

Temperature and Doppler Coefficients of Various Space Nuclear Reactors

Science.gov (United States)

Mughabghab, Said F.; Ludewig, Hans; Schmidt, Eldon

1994-07-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system.

Nuclear reactor installation

International Nuclear Information System (INIS)

Jungmann, A.

1976-01-01

A nuclear reactor metal pressure vessel is surrounded by a concrete wall forming an annular space around the vessel. Thermal insulation is in this space and surrounds the vessel, and a coolant-conductive layer is also in this space surrounding the thermal insulation, coolant forced through this layer reducing the thermal stress on the concrete wall. The coolant-conductive layer is formed by concrete blocks laid together and having coolant passages, these blocks being small enough individually to permit them to be cast from concrete at the reactor installation, the thermal insulation being formed by much larger sheet-metal clad concrete segments. Mortar is injected between the interfaces of the coolant-conductive layer and concrete wall and the interfaces between the fluid-conductive layer and the insulation, a layer of slippery sheet material being interposed between the insulation and the mortar. When the pressure vessel is thermally expanded by reactor operation, the annular space between it and the concrete wall is completely filled by these components so that zero-excursion rupture safeguard is provided for the vessel. 4 claims, 1 figure

Role of nuclear reactors in future military satellites

International Nuclear Information System (INIS)

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)/

The role of nuclear power and nuclear propulsion in the peaceful exploration of space

International Nuclear Information System (INIS)

2005-09-01

This publication has been produced within the framework of the IAEA's innovative reactor and fuel cycle technology development activities. It elucidates the role that peaceful space related nuclear power research and development could play in terrestrial innovative reactor and fuel cycle technology development initiatives. This review is a contribution to the Inter-Agency Meeting on Outer Space Activities, and reflects the stepped up efforts of the Scientific and Technical Subcommittee of the Committee on the Peaceful Uses of Outer Space to further strengthen cooperation between international organizations in space related activities. Apart from fostering information exchange within the United Nations organizations, this publication aims at finding new potential fields for innovative reactor and fuel cycle technology development. In assessing the status and reviewing the role of nuclear power in the peaceful exploration of space, it also aims to initiate a discussion on the potential benefits of space related nuclear power technology research and development to the development of innovative terrestrial nuclear systems

Space ecoliteracy- five informal education models for community empowerment

Science.gov (United States)

Venkataramaiah, Jagannatha; Jagannath, Sahana; J, Spandana; J, Sadhana; Jagannath, Shobha

Space ecoliteracy is a historical necessity and vital aspect of space age.Space Situational Awareness has taught lessons for mankind to look inward while stretching beyond cradle in human endeavours. Quality of life for every one on the only home of mankind-TERRA shall be a feasibility only after realizing Space ecoliteracy amongst all stakeholders in space quest. Objectives of Informal Environmental Education(UNESCO/UNEP/IEEP,1977) mandates awareness, attitude, knowledge, skill and participation at Individual and Community domains. Application of Space Technology at both Telecommunications and Remote Sensing domain have started making the fact that mankind has a challenge to learn and affirm earthmanship. Community empowerment focus after Earth Summit 1992 mandate of Sustainable Development has demonstrated a deluge of best practices in Agriculture,Urban, Industries and service sectors all over the globe. Further, deployment of Space technologies have proved the immense potential only after pre-empting the participatory approach at individual and community levels.Indian Space Programme with its 44th year of space service to national development has demonstrated self reliance in space technology for human development. Space technology for the most underdeveloped is a success story both in communication and information tools for quality of life. In this presentation Five Space Ecoliteracy models designed and validated since 1985 till date on informal environmental education namely 1) Ecological Environmental Studies by Students-EESS (1988): cited as one of the 20 best eco -education models by Earth Day Network,2)Community Eco Literacy Campaign-CEL,(2000): cited as a partner under Clean Up the World Campaign,UN, 3) Space Eco Literacy(2011)-an informa 8 week space eco literacy training reported at 39th COSPAR 12 assembly and 4) Space Eco Literacy by Practice(2014)- interface with formal education at institutions and 5) Space Ecoliteracy Mission as a space out reach in

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

Simplified methodology for control cell constant calculations of the reactor cores for the space kinetics

International Nuclear Information System (INIS)

Santos, Rubens Souza dos; Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques

2002-01-01

In this work is presented a methodology which focuses the distribution of neutron absorber rods in nuclear reactor power plants, for utilizing in space kinetic calculations, principally in the cluster ejection transients of control rods. A numerical model for macroscopic constant calculations based on the knowledge of the neutron flux without the control rods is proposed, as alternative to the analytical models, based on the hypothesis of the null current on the cell super boundaries. The proposed model in this work has itself showed adequate to deal with problems with strong space dependence, once that the model showed consistence in the global average built in the analytical model. (author)

Maps of Information Spaces: Assessments from Astronomy.

Science.gov (United States)

Poincot, Philippe; Lesteven, Soizick; Murtagh, Fionn

2000-01-01

Discusses the implementation of a cartographic user interface to bibliographic and other information sub-spaces in astronomy. Presents a range of comparative assessments, in operational frameworks, of this approach to accessing and retrieving astronomical information. Discusses the particular role that such cartographic user interfaces can play in…

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

International Nuclear Information System (INIS)

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

Value-informed space systems design and acquisition

Science.gov (United States)

Brathwaite, Joy

Investments in space systems are substantial, indivisible, and irreversible, characteristics that make them high-risk, especially when coupled with an uncertain demand environment. Traditional approaches to system design and acquisition, derived from a performance- or cost-centric mindset, incorporate little information about the spacecraft in relation to its environment and its value to its stakeholders. These traditional approaches, while appropriate in stable environments, are ill-suited for the current, distinctly uncertain, and rapidly changing technical and economic conditions; as such, they have to be revisited and adapted to the present context. This thesis proposes that in uncertain environments, decision-making with respect to space system design and acquisition should be value-based, or at a minimum value-informed. This research advances the value-centric paradigm by providing the theoretical basis, foundational frameworks, and supporting analytical tools for value assessment of priced and unpriced space systems. For priced systems, stochastic models of the market environment and financial models of stakeholder preferences are developed and integrated with a spacecraft-sizing tool to assess the system's net present value. The analytical framework is applied to a case study of a communications satellite, with market, financial, and technical data obtained from the satellite operator, Intelsat. The case study investigates the implications of the value-centric versus the cost-centric design and acquisition choices. Results identify the ways in which value-optimal spacecraft design choices are contingent on both technical and market conditions, and that larger spacecraft for example, which reap economies of scale benefits, as reflected by their decreasing cost-per-transponder, are not always the best (most valuable) choices. Market conditions and technical constraints for which convergence occurs between design choices under a cost-centric and a value

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1991-01-01

This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1989 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials. 11 figs., 3 tabs

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1992-06-01

The circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1991 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials

Colliding beam fusion reactor space propulsion system

International Nuclear Information System (INIS)

Wessel, Frank J.; Binderbauer, Michl W.; Rostoker, Norman; Rahman, Hafiz Ur; O'Toole, Joseph

2000-01-01

We describe a space propulsion system based on the Colliding Beam Fusion Reactor (CBFR). The CBFR is a high-beta, field-reversed, magnetic configuration with ion energies in the range of hundreds of keV. Repetitively-pulsed ion beams sustain the plasma distribution and provide current drive. The confinement physics is based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all sources and sinks for energy and particle flow. The mean azimuthal velocities and temperatures of the fuel ion species are equal and the plasma current is unneutralized by the electrons. The resulting distribution functions are thermal in a moving frame of reference. The ion gyro-orbit radius is comparable to the dimensions of the confinement system, hence classical transport of the particles and energy is expected and the device is scaleable. We have analyzed the design over a range of 10 6 -10 9 Watts of output power (0.15-150 Newtons thrust) with a specific impulse of, I sp ∼10 6 sec. A 50 MW propulsion system might involve the following parameters: 4-meters diameterx10-meters length, magnetic field ∼7 Tesla, ion beam current ∼10 A, and fuels of either D-He 3 ,P-B 11 ,P-Li 6 ,D-Li 6 , etc

Application of a Systems Engineering Approach to Support Space Reactor Development

International Nuclear Information System (INIS)

Wold, Scott

2005-01-01

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

Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

International Nuclear Information System (INIS)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung

2015-01-01

The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW th power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H 2 corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the preliminary design

Deciphering robust reactor kinetic data using mutual information

International Nuclear Information System (INIS)

Kumar, P.T. Krishna

2007-01-01

Experimentalists use Chauvenets's criterion to check the quality of any measured data. Based on this criterion they rejected data having high degree of correlation. Multivariate techniques like principal component analysis used for analysis of these correlated data, does not provide any scope to minimize the effect of correlation. We propose a novel method using information theory and the technique of determinant inequalities developed by us to reduce the effect of correlation among these data without summarily rejecting them. We demonstrate the utility of our technique in transient measurements of kinetic parameters performed on the commercially advanced gas cooled reactor (CAGCR)

10- to 30-kWe space power system using the uranium-zirconium hydride reactor and organic Rankine power conversion system

International Nuclear Information System (INIS)

Determan, W.R.; Bost, D.S.

1987-01-01

The UZrH reactor-ORC power system has been reviewed to determine its feasibility issues and characterize the system size, mass, and efficiency in the 10- to 30-kWe power range. The major component technologies required for this concept were reviewed to determine their technology status rating for early deployment of the system on near-term missions. Dynamic Isotope Power System (DIPS) technology is directly applicable to the UZrH reactor-ORC concept in the areas of power system reliability and survivability. The UZrH reactor-ORC concept provides a truly state-of-the-art system for use in future military and civilian space power programs. 9 references

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Opinions of the well-informed persons about the nuclear reactor facility periodical inspection

International Nuclear Information System (INIS)

Aeba, Yoichi; Ishikawa, Michio; Enomoto, Toshiaki; Oomori, Katsuyoshi

2005-01-01

Falsifications of self-inspection records in the shrouds and of leakage rates for containment vessels at TEPCO nuclear power plants destroyed public trust in nuclear safety. The Nuclear Reactor Regulation Law and Electric Utility Law were amended to enhance the nuclear safety regulation system. The major improvements are that operators are legally required to conduct inspection (periodical operator inspection) and recording and keeping inspection results. The operator performs 'periodical operator inspection' regularly, and Nuclear and Industrial Safety Agency (NISA) performs periodical inspection' about particularly important facilities/function in safety. Sixteen opinions of well-informed persons about the nuclear reactor facility periodical inspection were presented in this special number. Interval of periodical inspection less than thirteen months was disputed. Maintenance activities should be more rationalized based on risk information. (T. Tanaka)

Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

International Nuclear Information System (INIS)

Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

1994-01-01

Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

Ground test facility for nuclear testing of space reactor subsystems

International Nuclear Information System (INIS)

Quapp, W.J.; Watts, K.D.

1985-01-01

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Hanson, Duane J. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States)); Arcieri, William C. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States)); Ward, Leonard W. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States))

1994-07-01

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment. ((orig.))

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

International Nuclear Information System (INIS)

Hanson, Duane J.; Arcieri, William C.; Ward, Leonard W.

1994-01-01

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment. ((orig.))

The Design Space of Information Presentation

DEFF Research Database (Denmark)

May, Michael; Petersen, Johannes

2007-01-01

A semiotic approach to the design space of information presentation is presented in which Formal Concept Analysis (FCA) is used to represent and explore attributes of abstract sign types and the media (graphical, haptic, acoustic, gestic) through which they are presented as specific...

The Orbital Space Environment and Space Situational Awareness Domain Ontology - Toward an International Information System for Space Data

Science.gov (United States)

Rovetto, R.

2016-09-01

The orbital space environment is home to natural and artificial satellites, debris, and space weather phenomena. As the population of orbital objects grows so do the potential hazards to astronauts, space infrastructure and spaceflight capability. Orbital debris, in particular, is a universal concern. This and other hazards can be minimized by improving global space situational awareness (SSA). By sharing more data and increasing observational coverage of the space environment we stand to achieve that goal, thereby making spaceflight safer and expanding our knowledge of near-Earth space. To facilitate data-sharing interoperability among distinct orbital debris and space object catalogs, and SSA information systems, I proposed ontology in (Rovetto, 2015) and (Rovetto and Kelso, 2016). I continue this effort toward formal representations and models of the overall domain that may serve to improve peaceful SSA and increase our scientific knowledge. This paper explains the project concept introduced in those publications, summarizing efforts to date as well as the research field of ontology development and engineering. I describe concepts for an ontological framework for the orbital space environment, near-Earth space environment and SSA domain. An ontological framework is conceived as a part of a potential international information system. The purpose of such a system is to consolidate, analyze and reason over various sources and types of orbital and SSA data toward the mutually beneficial goals of safer space navigation and scientific research. Recent internationals findings on the limitations of orbital data, in addition to existing publications on collaborative SSA, demonstrate both the overlap with this project and the need for datasharing and integration.

Nuclear power in space

International Nuclear Information System (INIS)

Anghaie, S.

2007-01-01

The development of space nuclear power and propulsion in the United States started in 1955 with the initiation of the ROVER project. The first step in the ROVER program was the KIWI project that included the development and testing of 8 non-flyable ultrahigh temperature nuclear test reactors during 1955-1964. The KIWI project was precursor to the PHOEBUS carbon-based fuel reactor project that resulted in ground testing of three high power reactors during 1965-1968 with the last reactor operated at 4,100 MW. During the same time period a parallel program was pursued to develop a nuclear thermal rocket based on cermet fuel technology. The third component of the ROVER program was the Nuclear Engine for Rocket Vehicle Applications (NERVA) that was initiated in 1961 with the primary goal of designing the first generation of nuclear rocket engine based on the KIWI project experience. The fourth component of the ROVER program was the Reactor In-Flight Test (RIFT) project that was intended to design, fabricate, and flight test a NERVA powered upper stage engine for the Saturn-class lunch vehicle. During the ROVER program era, the Unites States ventured in a comprehensive space nuclear program that included design and testing of several compact reactors and space suitable power conversion systems, and the development of a few light weight heat rejection systems. Contrary to its sister ROVER program, the space nuclear power program resulted in the first ever deployment and in-space operation of the nuclear powered SNAP-10A in 1965. The USSR space nuclear program started in early 70's and resulted in deployment of two 6 kWe TOPAZ reactors into space and ground testing of the prototype of a relatively small nuclear rocket engine in 1984. The US ambition for the development and deployment of space nuclear powered systems was resurrected in mid 1980's and intermittently continued to date with the initiation of several research programs that included the SP-100, Space Exploration

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

Content Sharing Based on Personal Information in Virtually Secured Space

Science.gov (United States)

Sohn, Hosik; Ro, Yong Man; Plataniotis, Kostantinos N.

User generated contents (UGC) are shared in an open space like social media where users can upload and consume contents freely. Since the access of contents is not restricted, the contents could be delivered to unwanted users or misused sometimes. In this paper, we propose a method for sharing UGCs securely based on the personal information of users. With the proposed method, virtual secure space is created for contents delivery. The virtual secure space allows UGC creator to deliver contents to users who have similar personal information and they can consume the contents without any leakage of personal information. In order to verify the usefulness of the proposed method, the experiment was performed where the content was encrypted with personal information of creator, and users with similar personal information have decrypted and consumed the contents. The results showed that UGCs were securely shared among users who have similar personal information.

Proceedings of the US Nuclear Regulatory Commission nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-04-01

This three-volume report contains 83 papers out of the 108 that were presented at the Nineteenth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 28--30, 1991. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 14 different papers presented by researchers from Canada, Germany, France, Japan, Sweden, Taiwan, and USSR. This document, Volume 3, presents papers on: Structural engineering; Advanced reactor research; Advanced passive reactors; Human factors research; Human factors issues related to advanced passive light water researchers; Thermal Hydraulics; and Earth sciences. The individual papers have been cataloged separately

SPACE 365: Upgraded App for Aviation and Space-Related Information and Program Planning

Science.gov (United States)

Williams, S.; Maples, J. E.; Castle, C. E.

2014-12-01

Foreknowledge of upcoming events and anniversary dates can be extraordinarily valuable in the planning and preparation of a variety of aviation and Space-related educational programming. Alignment of programming with items "newsworthy" enough to attract media attention on their own can result in effective program promotion at low/no cost. Similarly, awareness and avoidance of dates upon which media and public attention will likely be elsewhere can keep programs from being lost in the noise.NASA has created a useful and entertaining app called "SPACE 365" to help supply that foreknowledge. The app contains an extensive database of historical aviation and Space exploration-related events, along with other events and birthdays to provide socio-historical context, as well as an extensive file of present and future space missions, complete with images and videos. The user can search by entry topic category, date, and key words. Upcoming Events allows the user to plan, participate, and engage in significant "don't miss" happenings.The historical database was originally developed for use at the National Air and Space Museum, then expanded significantly to include more NASA-related information. The CIMA team at NASA MSFC, sponsored by the Planetary Science Division, added NASA current events and NASA educational programming information, and are continually adding new information and improving the functionality and features of the app. Features of SPACE 365 now include: NASA Image of the Day, Upcoming NASA Events, Event Save, Do Not Miss, and Ask Dr. Steve functions, and the CIMA team recently added a new start page and added improved search and navigation capabilities. App users can now socialize the Images of the Day via Twitter, Pinterest, Facebook, and other social media outlets.SPACE 365 is available at no cost from both the Apple appstore and GooglePlay, and has helped NASA, NASM, and other educators plan and schedule programming events. It could help you, too!

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

Research on the method of measuring space information network capacity in communication service

Directory of Open Access Journals (Sweden)

Zhu Shichao

2017-02-01

Full Text Available Because of the large scale characteristic of space information network in terms of space and time and the increasing of its complexity,existing measuring methods of information transmission capacity have been unable to measure the existing and future space information networkeffectively.In this study,we firstly established a complex model of space information network,and measured the whole space information network capacity by means of analyzing data access capability to the network and data transmission capability within the network.At last,we verified the rationality of the proposed measuring method by using STK and Matlab simulation software for collaborative simulation.

Space Sharing Between Formal and Informal Sectors In Kemang Area

Science.gov (United States)

Sihombing, Antony; Dewanti, Hafizka Chandra

2018-01-01

Sharing is a form of human activity as a social being, over resources or spaces. Humans generally define their space according to their psychological and biological needs. However, what if space sharing takes place in an urban scope? The phenomenon of space sharing happens in Kemang area done by two distinct sectors, formal and informal which both are engaged in commercial activities independently. In the study of territory and the behavior settings, the quality of environment can affect the formation of human activities in a space, occurs a phenomenon of space sharing. The researcher will discuss and present a form of the space sharing by the formal and informal sectors in a commercial area through different environmental qualities. In some circumstance, a form of space sharing can increase the value of space and obtain to a shared space, where both sectors mutually take benefits. Otherwise, it leads to an adverse impact in some others. It is clear that differences in the physical environment and social environment have an impact on the formation of space sharing and the value of space in an urban region.

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

The Paucity Problem: Where Have All the Space Reactor Experiments Gone?

Energy Technology Data Exchange (ETDEWEB)

Bess, John D.; Marshall, Margaret A.

2016-10-01

The Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) together contain a plethora of documented and evaluated experiments essential in the validation of nuclear data, neutronics codes, and modeling of various nuclear systems. Unfortunately, only a minute selection of handbook data (twelve evaluations) are of actual experimental facilities and mockups designed specifically for space nuclear research. There is a paucity problem, such that the multitude of space nuclear experimental activities performed in the past several decades have yet to be recovered and made available in such detail that the international community could benefit from these valuable historical research efforts. Those experiments represent extensive investments in infrastructure, expertise, and cost, as well as constitute significantly valuable resources of data supporting past, present, and future research activities. The ICSBEP and IRPhEP were established to identify and verify comprehensive sets of benchmark data; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data. See full abstract in attached document.

Status of CEA reactor studies for a 200 kWe turbo electric space power system

International Nuclear Information System (INIS)

Carre, F.; Gervaise, F.; Proust, E.; Schwartz, J.P.; Tilliette, Z.; Vrillon, B.

1986-01-01

The present European ARIANE space program will expand after 1995 in the development of the large ARIANE 5 launch vehicle. Considering, that the range of power needs (50 to 400 kWe) and operation times required for the space missions planned after the year 2000, are relevant to a nuclear power system, the French Centre National d'Etudes Spatiales (CNES) invited in 1983 the Commissariat a l'Energie Atomique (CEA) to undertake preliminary studies on space power systems. The purpose of the present two year phase (mid 1984-mid 1986) is to identify key technologies for a space generator within the power range of interest and to estimate the development cost of such a project to be examined for commitment in 1986. This work mainly consists in the feasibility and cost assessment of a reference 200 kWe turboelectric space generator, selected for the maturity and availability of the conversion system and for its attractive specific mass compared to thermionics and thermoelectricity, considering the available radiator area afforded by the specific ARIANE 5 geometrical features. The system is basically composed of a fast neutron spectrum lithium cooled reactor, of a Brayton conversion loop and of a heat pipe radiator

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1993-03-01

This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96--295. The report provides a brief description of NRC authority for certain aspects of transporting spent fuel. It provides descriptive statistics on spent fuel shipments regulated by the NRC from 1979 to 1992. It also lists detailed highway and railway segments used within each state from October 1, 1987 through December 31, 1992

Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

2015-10-15

The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW{sub th} power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H{sub 2} corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Modified parity space averaging approaches for online cross-calibration of redundant sensors in nuclear reactors

Directory of Open Access Journals (Sweden)

Moath Kassim

2018-05-01

Full Text Available To maintain safety and reliability of reactors, redundant sensors are usually used to measure critical variables and estimate their averaged time-dependency. Nonhealthy sensors can badly influence the estimation result of the process variable. Since online condition monitoring was introduced, the online cross-calibration method has been widely used to detect any anomaly of sensor readings among the redundant group. The cross-calibration method has four main averaging techniques: simple averaging, band averaging, weighted averaging, and parity space averaging (PSA. PSA is used to weigh redundant signals based on their error bounds and their band consistency. Using the consistency weighting factor (C, PSA assigns more weight to consistent signals that have shared bands, based on how many bands they share, and gives inconsistent signals of very low weight. In this article, three approaches are introduced for improving the PSA technique: the first is to add another consistency factor, so called trend consistency (TC, to include a consideration of the preserving of any characteristic edge that reflects the behavior of equipment/component measured by the process parameter; the second approach proposes replacing the error bound/accuracy based weighting factor (Wa with a weighting factor based on the Euclidean distance (Wd, and the third approach proposes applying Wd,TC,andC, all together. Cold neutron source data sets of four redundant hydrogen pressure transmitters from a research reactor were used to perform the validation and verification. Results showed that the second and third modified approaches lead to reasonable improvement of the PSA technique. All approaches implemented in this study were similar in that they have the capability to (1 identify and isolate a drifted sensor that should undergo calibration, (2 identify a faulty sensor/s due to long and continuous missing data range, and (3 identify a healthy sensor. Keywords: Nuclear Reactors

Space dynamics

International Nuclear Information System (INIS)

Corno, S.E.

1995-01-01

Analytical methods for Space Dynamics of fission reactors, are presented. It is shown how a few sample problems in space dynamics can be solved, within the one and two group diffusion model, by purely analytical tools, essentially based on Laplace transform and complex Green function techniques. A quite suggestive generalization of this approach, applicable to the fluid core reactors, whose fuel is undergoing a violent mixing, is reported and briefly discussed. (author)

Ground testing of an SP-100 prototypic reactor

International Nuclear Information System (INIS)

Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

1988-01-01

SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1994-04-01

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database

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

Science.gov (United States)

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

1977-01-01

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

RESI-1 and RESI-2: pPrototypes of an information system on reactor safety

International Nuclear Information System (INIS)

Schultheiss, G.F.; Eglin, W.; Katz, F.W.; Krings, T.; Pee, A.; Schlechtendahl, E.G.

1975-04-01

To demonstrate by practical experience the feasibility of the information system elaborated in the 'Study of an Information System on Reactor Safety RESI' (KFK 1900), the prototype systems RESI-1 and RESI-2 were developed and tested in operation. The two systems have been considerably reduced both in extent and contents as compared to the information system described in the study. The RESI-1 prototype system is a paper version established for verification of all the individual functions before passing over to the computer-aided interactive version RESI-2. RESI-2 is based on the GOLEM system of Siemens. Both protoype systems have proved that the essential features: 1) documentation, 2) formulation of and answering to safety questions, which are relevant with respect to particular licensing cases, 3) formulation of safety questions related to individual reactor types can be managed satisfactorily. All the functions of information retrieval have been tested carefully over several months. Particularities of project development and of the methods elaborated are described in detail and presented in this report. (orig.) [de

Demonstration and information center on the basis of the research reactor IR-50

International Nuclear Information System (INIS)

Krupenina, F.

2001-01-01

Many problems exist in the nuclear field, but the most significant one is the public's mistrust of Nuclear Energy. Strong downfalls of the radiological culture affect public perception, the main paradox being the situation after Chernobyl. The task of creating a Demonstration and-Information Center (Minatom RF) on the basis of the research reactor IR-50 is conducted by Research and Development Institute of Power Engineering (ENTEK). The IR-50 is situated on the grounds of the institute. It will be a unique event when the functional reactor is situated in the center of the city (about 5 km from Kremlin). (author)

Nuclear reactor installation

International Nuclear Information System (INIS)

Keller, W.

1976-01-01

A nuclear reactor installation includes a pressurized-water coolant reactor vessel and a concrete biological shield surrounding this vessel. The shield forms a space between it and the vessel large enough to permit rapid escape of the pressurized-water coolant therefrom in the event the vessel ruptures. Struts extend radially between the vessel and shield for a distance permitting normal radial thermal movement of the vessel, while containing the vessel in the event it ruptures, the struts being interspaced from each other to permit rapid escape of the pressurized-water coolant from the space between the shield and the vessel

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

Nuclear power in space

International Nuclear Information System (INIS)

Aftergood, S.; Hafemeister, D.W.; Prilutsky, O.F.; Rodionov, S.N.; Primack, J.R.

1991-01-01

Nuclear reactors have provided energy for satellites-with nearly disastrous results. Now the US government is proposing to build nuclear-powered boosters to launch Star Wars defenses. These authors represent scientific groups that are opposed to the use of nuclear power in near space. The authors feel that the best course for space-borne reactors is to ban them from Earth orbit and use them in deep space

Reactor container

International Nuclear Information System (INIS)

Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

1990-01-01

In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

Light-water reactors: preliminary safety and environmental information document. Volume I

International Nuclear Information System (INIS)

1980-01-01

Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle

The Information Infrastructures Design Space

DEFF Research Database (Denmark)

Henningsson, Stefan; Rapti, Charikleia; Jensen, Thomas

2017-01-01

This paper develops a framework for characterising the design space of Information Infrastructures (IIs). Existing research has generally sought to unravel the convergent characteristics and mechanisms uniting IIs across a wide range of manifestations. In this research, we explore this divergence...... within the II design space. We do so by reviewing the II literature, focusing on the two domains of design situation and design resolution. Design situation refers to the relevant dimensions of the context in which an II is employed. Design resolution covers the dimensions along which the socio......-technical constituents can be assembled to form an effective solution. The resulting framework allows for the comparing and contrasting of II initiatives, and contributes towards a cumulative knowledge process aimed at a more refined understanding of how an II can be configured to address the specific problem at hand....

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

Meeting the reactor operator's information needs using functional analysis

Energy Technology Data Exchange (ETDEWEB)

Nelson, W.R.; Clark, M.T.

1980-01-01

Since the accident at Three Mile Island, many ideas have been proposed for assisting the reactor operator during emergency situations. However, some of the suggested remedies do not alleviate an important shortcoming of the TMI control room: the operators were not presented with the information they needed in a manner which would allow prompt diagnosis of the problem. To address this problem, functional analysis is being applied at the LOFT facility to ensure that the operator's information needs are being met in his procedures and graphic displays. This paper summarizes the current applications of functional analysis at LOFT.

Methodology of nuclear reactor monitoring and diagnostics using information dimension

International Nuclear Information System (INIS)

Suzudo, Tomoaki; Hayashi, Koji; Shinohara, Yoshikuni

1993-01-01

Reactor noise analysis method based on information dimension is applied to the monitoring and diagnosing of power oscillation. The method focuses on the utilization of the slope of the correlation integral (SOCI) which determines the information dimension of attractors. For practical application, the information dimension is expected to be the same as the fractal dimension of attractors; it can be used to classify different asymptotic regimes of nonlinear dynamical systems. We examined a real power oscillation using SOCI and the results implied that the oscillation was just a noisy limit cycle, although it is not possible to assert that there is no chaotic character in the oscillation because large oscillatory time-series data sets are not available. In addition, the application of SOCI to the real-time monitoring of power oscillation is proposed and examined. (author)

Development and computational simulation of thermoelectric electromagnetic pumps for controlling the fluid flow in liquid metal cooled space nuclear reactors

International Nuclear Information System (INIS)

Borges, E.M.

1991-01-01

Thermoelectric Electromagnetic (TEEM) Pumps can be used for controlling the fluid flow in the primary and secondary circuits of liquid metal cooled space nuclear reactor. In order to simulate and to evaluate the pumps performance, in steady-state, the computer program BEMTE has been developed to study the main operational parameters and to determine the system actuation point, for a given reactor operating power. The results for each stage of the program were satisfactory, compared to experimental data. The program shows to be adequate for the design and simulating of direct current electromagnetic pumps. (author)

Mathematical foundation of the application of modal analysis to the investigation of space-time reactor behaviour

International Nuclear Information System (INIS)

Obradovic, D.M.

1970-01-01

In recent years investigations in the field of kinetics and dynamics of nuclear reactors have been directed towards overcoming an insufficiently accurate point reactor model. For that purpose different mathematical approaches have been used. This thesis is devoted to modal analysis because, from the practical point of view, it is a very promising and, from the mathematical and physical point of view, a very interesting method. Some fundamental mathematical problems connected with the application of modal analysis to the investigations of the reactor space-time behaviour are still unsolved and accordingly our purpose is to solve some of these problems. The spectral properties of the diffusion and P 1 operators are studied in some detail applying the Krein-Rutman theory of the K-positive operators, the Krasnosel'skii theory of u 0 operators, and the Keldis theory of the operator families. The formal solution to the initial value problem (as an abstract Cauchy problem), associated with the diffusion and P 1 operators is also studied. Modal analysis is identified as a set of methods in the mathematical literature known as the Galerkin methods (or projection methods). Following this idea (using the results of the mathematical investigations of the Galerkin methods) and using our results of the investigations of the properties of the diffusion and P 1 operators, the applicability of modal analysis to the approximate solution of the diffusion and P 1 equations and of the eigenvalue problems associated with the diffusion and P 1 operators is established. As an example of the application of modal analysis the Bubnov and Galerkin method is applied to a multiregion thermal nuclear reactor for the determination of: (i) frequency response, (ii) eigenvalues and eigenvectors of the stationary diffusion operator, (iii) eigenvalues and eigenvectors of the non-stationary diffusion operators. On the basis of the expressions obtained the corresponding computer programmes for radial

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1988-01-01

According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

Space nuclear reactor electric power. January 1972-May 1990 (A Bibliography from the International Aerospace Abstracts data base). Report for January 1972-May 1990

International Nuclear Information System (INIS)

1990-05-01

This bibliography contains citations concerning studies and conceptual designs of nuclear space power reactors to generate electric power for space missions. The citations cover the technology, safety aspects, and policy considerations. (This updated bibliography contains 280 citations, 172 of which are new entries to the previous edition.)

Nuclear reactor vessel fuel thermal insulating barrier

Science.gov (United States)

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.

Urban Public Space Context and Cognitive Psychology Evolution in Information Environment

Science.gov (United States)

Feng, Chen; Xu, Hua-wei

2017-11-01

The rapid development of information technology has had a great impact on the understanding of urban environment, which brings different spatially psychological experience. Information and image transmission has been full with the streets, both the physical space and virtual space have been unprecedentedly blended together through pictures, images, electronic media and other tools, which also stimulates people’s vision and psychology and gives birth to a more complex form of urban space. Under the dual role of spatial mediumlization and media spatialization, the psychological cognitive pattern of urban public space context is changing.

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

Summary of particle bed reactor designs for the Space Nuclear Thermal Propulsion Program

Science.gov (United States)

Powell, J. R.; Ludewig, H.; Todosow, M.

1993-09-01

A summary report of the Particle Bed Reactor (PBR) designs considered for the space nuclear thermal propulsion program has been prepared. The first chapters outline the methods of analysis, and their validation. Monte Carlo methods are used for the physics analysis, several new algorithms are used for the fluid dynamics heat transfer and engine system analysis, and commercially available codes are used for the stress analysis. A critical experiment, prototypic of the PBR was used for the physics validation, and blowdown experiments using fuel beds of prototypic dimensions were used to validate the power extraction capabilities from particle beds. In all four different PBR rocket reactor designs were studied to varying degrees of detail. They varied in power from 400 MW to 2000 MW. These designs were all characterized by a negative prompt coefficient, due to Doppler feedback, and the feedback due to moderator heat up varied from slightly negative to slightly positive. In all practical cases, the coolant worth was positive, although core configurations with negative coolant worth could be designed. In all practical cases the thrust/weight ratio was greater than 20.

Small power reactor projects in the United States of America and Canada. Information gathered as a result of invitations from Member States

International Nuclear Information System (INIS)

1962-01-01

As part of its activities in connection with the development of nuclear power, and in response to the resolutions adopted by the General Conference, the Agency has been undertaking a continuing study of the technology and economics of small and medium sized power reactors, particularly with reference to the needs of the less-developed countries. This report summarizes the information gathered on the small power reactor projects in the United States of America and Canada, as a result of the opportunity afforded by these Member States to the Agency. It may be recalled that, at the third regular session of the General Conference, the United States Government offered to provide the Agency with relevant technical and economic data on several small power reactor projects of its Atomic Energy Commission. The Agency accepted the offer and since June 1960 it has sent one or two staff members at approximately six-monthly intervals to follow the development of nine power reactor projects in the United States which represent six different reactor systems. Last year, the Agency issued a report summarizing the information obtained through their visits and study of available published literature. The present document, which should be read in conjunction with that document, brings the information up to date and provides additional information on certain phases of the projects already discussed in the last report. Three more power reactor projects are also dealt with, namely the experimental gas-cooled reactor (EGCR), the high temperature gas-cooled reactor (HTGR) and the Hallam nuclear power facility (HNPF). Early in 1962, the Canadian Government expressed its willingness to make available to the Agency relevant information on the NPD and CANDU projects. The coverage of the NPD reactor is based upon the published information supplied by AECL of Canada and the visit by one of the staff members to the NPD site. The Agency wishes to acknowledge with thanks the co-operation extended

Facility with a nuclear district heating reactor

International Nuclear Information System (INIS)

Straub, H.

1988-01-01

The district heating reactor has a pressure vessel which contains the reactor core and at least one coolant conducting primary heat carrier surrounded by a heat sink. The pressure vessel has two walls with a space between them. This space is connected with a container which contains air as heat isolating medium and water as heat conducting medium. During the normal reactor operation the space is filled by air from the container with the aid of a blower, whereas in the case of a break-down of the cooling system it is filled by water which flows out of the container by gravity after the blower has been switched off. The after-heat, generated in the reactor core during cooling break-down, is removed into the heat sink surrounding the pressure vessel in a safe and simple way. 6 figs

Mechanical characterization of superalloys for space reactors

International Nuclear Information System (INIS)

Duchesne, J.

1989-01-01

The purpose of this work is the choice of materials usable between 600 and 900 0 C for nuclear space reactor structures. The main criterion of selection for these materials is their good creep behaviour. Consequently, macroscopic theories of creep and several extrapolation methods were described. Superalloys seem the best materials for the studied range of temperatures. Five of them, base nickel, ones unusual in nuclear industry were selected for their good mechanical properties. Three of them are industrial alloys: the first, HAYNES 230 is a recent one, HASTELLOY S and X are more standard materials. The last two, HASTELLOY XR and PYRAD 38 D are issued from special fabrications. Creep tests metallographic investigations, hardness and tensile tests were performed. A contraction of samples was observed during some creep tests under a low stress, 20MPa at 800 0 C, for HAYNES 230 and HASTELLOY X. This could be due to a structural evolution of these materials connected to a decrease of the cristalline parameter. In addition, correlations were observed between certain characteristics determined from slow tensile tests and short duration creep tests. These correlations present a large interest because, at the present time, creep tests cannot be executed on irradiated materials in our laboratories. Consequently creep behaviour of irradiated materials seem may be deduced. Further studies are needed to explain and confirm the behaviour of the most interesting materials under low stresses: HAYNES 230 and HASTELLOY XR to anticipate their behaviour in working conditions [fr

Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

Energy Technology Data Exchange (ETDEWEB)

Montes, J

1972-07-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Principles elaboration and creation of information-analytical system ''RI Operation Safety with SSC RIAR Research Reactors''

International Nuclear Information System (INIS)

Ivanov, V.B.; Grachev, A.F.; Kinsky, O.M.; Makin, P.S.; Okhrimenko, A.I.; Demidov, L.I.; Karpyuk, V.I.; Afonin, V.K.; Iskanderov, R.G.

1995-01-01

In this paper an approach is described, which is accepted at elaboration and creation of computer-aided control system of technological process (CCS TP) at the installations with research reactors. The tasks and the main technological requirements to elaborated information-analytical system, are formulated, based on the accepted approach, experience of computer-aided systems and analysis of technological processes at reactor installations (RI) of SSC RIAR. The system includes the following installations: the SM-3, the VK-50, the RBT-10, the BOR-60 and the MIR. Based on the given example there is a classification and the purposes of the modern system of information personnel support of research reactors are formulated as well as approaches to its creation, including creation of determined models of the processes, which are realized in simulators and statistic methods of time series. According to the accepted approaches the results of systematic-technical synthesis and modern states with system simulation are described. (author). 17 refs, 3 figs, 4 tabs

The Information Science Experiment System - The computer for science experiments in space

Science.gov (United States)

Foudriat, Edwin C.; Husson, Charles

1989-01-01

The concept of the Information Science Experiment System (ISES), potential experiments, and system requirements are reviewed. The ISES is conceived as a computer resource in space whose aim is to assist computer, earth, and space science experiments, to develop and demonstrate new information processing concepts, and to provide an experiment base for developing new information technology for use in space systems. The discussion covers system hardware and architecture, operating system software, the user interface, and the ground communication link.

Space nuclear reactor power plants

International Nuclear Information System (INIS)

Buden, D.; Ranken, W.A.; Koenig, D.R.

1980-01-01

Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

Phase space view of quantum mechanical systems and Fisher information

Energy Technology Data Exchange (ETDEWEB)

Nagy, Á., E-mail: anagy@madget.atomki.hu

2016-06-17

Highlights: • Phase-space Fisher information coming from the canonical distribution is derived for the ground state of quantum mechanical systems. • Quantum mechanical phase-space Fisher information contains an extra term due to the position dependence of the temperature. • A complete analogy to the classical case is demonstrated for the linear harmonic oscillator. - Abstract: Pennini and Plastino showed that the form of the Fisher information generated by the canonical distribution function reflects the intrinsic structure of classical mechanics. Now, a quantum mechanical generalization of the Pennini–Plastino theory is presented based on the thermodynamical transcription of the density functional theory. Comparing to the classical case, the phase-space Fisher information contains an extra term due to the position dependence of the temperature. However, for the special case of constant temperature, the expression derived bears resemblance to the classical one. A complete analogy to the classical case is demonstrated for the linear harmonic oscillator.

Phase space view of quantum mechanical systems and Fisher information

International Nuclear Information System (INIS)

Nagy, Á.

2016-01-01

Highlights: • Phase-space Fisher information coming from the canonical distribution is derived for the ground state of quantum mechanical systems. • Quantum mechanical phase-space Fisher information contains an extra term due to the position dependence of the temperature. • A complete analogy to the classical case is demonstrated for the linear harmonic oscillator. - Abstract: Pennini and Plastino showed that the form of the Fisher information generated by the canonical distribution function reflects the intrinsic structure of classical mechanics. Now, a quantum mechanical generalization of the Pennini–Plastino theory is presented based on the thermodynamical transcription of the density functional theory. Comparing to the classical case, the phase-space Fisher information contains an extra term due to the position dependence of the temperature. However, for the special case of constant temperature, the expression derived bears resemblance to the classical one. A complete analogy to the classical case is demonstrated for the linear harmonic oscillator.

Ethical Issues Regarding Informed Consent for Minors for Space Tourism

Science.gov (United States)

Marsh, Melvin S.

2010-01-01

This paper describes the difficulty with informed consent and debates whether or not whether adults should be able to ethically, morally, and legally consent for their children during the high-risk activity of space tourism. The experimental nature of space vehicles combined with the high likelihood of medical complications and the destination places space tourism legally in the category of "adventure activities," which include adventure travel to exotic locations as well as adventure sports, such as mountain climbing, rafting, etc. which carry a high risk of danger (http://rescommunis.wordpress.com/2008/02/14/interview-tracey-l-knutson-adventure-sports-defense-attorney-on-space-tourism-risk-and-informed-consente/). However, unlike other adventure sports, adults currently cannot consent for their minor children. Other topics also receive attention, such as a "mature minors" clause, radiation exposure of potential future children, and other difficulties preventing adults from legally consenting to space travel.

Nuclear reactor theory

International Nuclear Information System (INIS)

Sekimoto, Hiroshi

2007-09-01

This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

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

Science.gov (United States)

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

Cermet fuel reactors

International Nuclear Information System (INIS)

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

Licensed operating reactors

International Nuclear Information System (INIS)

1989-08-01

THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

Licensed operating reactors

International Nuclear Information System (INIS)

1990-04-01

The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

Visual Navigation of Complex Information Spaces

Directory of Open Access Journals (Sweden)

Sarah North

1995-11-01

Full Text Available The authors lay the foundation for the introduction of visual navigation aid to assist computer users in direct manipulation of the complex information spaces. By exploring present research on scientific data visualisation and creating a case for improved information visualisation tools, they introduce the design of an improved information visualisation interface utilizing dynamic slider, called Visual-X, incorporating icons with bindable attributes (glyphs. Exploring the improvement that these data visualisations, make to a computing environment, the authors conduct an experiment to compare the performance of subjects who use traditional interfaces and Visual-X. Methodology is presented and conclusions reveal that the use of Visual-X appears to be a promising approach in providing users with a navigation tool that does not overload their cognitive processes.

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

Science.gov (United States)

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.

Yahoo! Answers as a Space for Informal Language Learning

Directory of Open Access Journals (Sweden)

Giuliana Dettori

2014-10-01

Full Text Available Online social spaces, where users can exchange information, opinions and resources, have achieved wide popularity and are gaining attention in many research fields, including education. Their actual potential support to learning, however, still requires investigation, especially because portals can widely differ as concerns purpose and internal structure. This paper aims to contribute in this respect, by concentrating on question answering, a kind of social space not yet widely discussed in education. We analyzed a small corpus of posts from the Languages section of Yahoo! Answers Italy, checking if the questions reveal some inclination to learning or just the desire to obtain a service and if the answers provided by the community members can be considered as reliable sources of knowledge. Our analysis highlights the presence of a variety of question/answer types, from mere information exchange or help for task completion, up to language-related questions prompting valuable short lessons. The quality of answers may widely vary as concerns pertinence, correctness and richness of supporting elements. We found a high number of purely task-oriented questions and answers, but also a higher number of learning-oriented questions and correct, informative answers. This suggests that this kind of social space actually has valuable potential for informal learning.

Broad-Application Test Reactor

International Nuclear Information System (INIS)

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

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

Science.gov (United States)

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

1993-01-01

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

Gene expression from polynomial dynamics in the 2-adic information space

International Nuclear Information System (INIS)

Khrennikov, Andrei Yu.

2009-01-01

We perform geometrization of genetics by representing genetic information by points of the 4-adic information space. By well known theorem of number theory this space can also be represented as the 2-adic space. The process of DNA-reproduction is described by the action of a 4-adic (or equivalently 2-adic) dynamical system. As we know, the genes contain information for production of proteins. The genetic code is a degenerate map of codons to proteins. We model this map as functioning of a polynomial dynamical system. The purely mathematical problem under consideration is to find a dynamical system reproducing the degenerate structure of the genetic code. We present one of possible solutions of this problem.

Bioconversion reactor

Science.gov (United States)

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.

Developments in modelling the effect of aerosol on the thermal performance of the Fast Reactor cover gas space

International Nuclear Information System (INIS)

Ford, I.J.; Clement, C.F.

1990-03-01

The sodium aerosol which forms in the cover gas space of a Fast Reactor couples the processes of heat and mass transfer to and from the bounding surfaces and affects the thermal performance of the cavity. This report describes extensions to previously separate models of heat transfer and aerosol formation and removal in the cover gas space, and the linking of the two calculations in a consistent manner. The extensions made to the theories include thermophoretic aerosol removal, radiative-driven redistribution in aerosol sizes, and the side-wall influence on the bulk cavity temperature. The link between aerosol properties and boundary layer saturations is also examined, especially in the far-from-saturated limit. The models can be used in the interpretation of cover gas space experiments and some example calculations are given. (author)

Information management system: A summary discussion. [for use in the space shuttle sortie, modular space station and TDR satellite

Science.gov (United States)

Sayers, R. S.

1972-01-01

An information management system is proposed for use in the space shuttle sortie, the modular space station, the tracking data relay satellite and associated ground support systems. Several different information management functions, including data acquisition, transfer, storage, processing, control and display are integrated in the system.

Public information circular for shipments of irradiated reactor fuel. Revision 12

International Nuclear Information System (INIS)

1997-10-01

This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1996 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials

Public information circular for shipments of irradiated reactor fuel. Revision 10

International Nuclear Information System (INIS)

1995-04-01

This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1994 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials

Parametric study of sodium aerosols in the cover-gas space of sodium-cooled reactors

International Nuclear Information System (INIS)

Sheth, A.

1975-03-01

A mathematical model has been developed to describe the behavior of sodium aerosols in the cover-gas space of a sodium-cooled reactor. A review of the literature was first made to examine methods of aerosol generation, mathematical expressions representing aerosol behavior, and pertinent experimental investigations of sodium aerosols. In the development of the model, some terms were derived from basic principles and other terms were estimated from available correlations. The model was simulated on a computer, and important parameters were studied to determine their effects on the overall behavior of sodium aerosols. The parameters studied were sodium pool temperature, source and initial size of particles, film thickness at the sodium pool/cover gas interface, wall plating parameters, cover-gas flow rate, and type of cover gas (argon and helium). The model satisfactorily describes the behavior of sodium aerosol in argon, but not in helium. Possible reasons are given for the failure of the model with helium, and further experimental work is recommended. The mathematical model, with appropriate modifications to describe the behavior of sodium aerosols in helium, would be very useful in designing traps to remove aerosols from the cover gas of sodium-cooled reactors. (U.S.)

Data Model Management for Space Information Systems

Science.gov (United States)

Hughes, J. Steven; Crichton, Daniel J.; Ramirez, Paul; Mattmann, chris

2006-01-01

The Reference Architecture for Space Information Management (RASIM) suggests the separation of the data model from software components to promote the development of flexible information management systems. RASIM allows the data model to evolve independently from the software components and results in a robust implementation that remains viable as the domain changes. However, the development and management of data models within RASIM are difficult and time consuming tasks involving the choice of a notation, the capture of the model, its validation for consistency, and the export of the model for implementation. Current limitations to this approach include the lack of ability to capture comprehensive domain knowledge, the loss of significant modeling information during implementation, the lack of model visualization and documentation capabilities, and exports being limited to one or two schema types. The advent of the Semantic Web and its demand for sophisticated data models has addressed this situation by providing a new level of data model management in the form of ontology tools. In this paper we describe the use of a representative ontology tool to capture and manage a data model for a space information system. The resulting ontology is implementation independent. Novel on-line visualization and documentation capabilities are available automatically, and the ability to export to various schemas can be added through tool plug-ins. In addition, the ingestion of data instances into the ontology allows validation of the ontology and results in a domain knowledge base. Semantic browsers are easily configured for the knowledge base. For example the export of the knowledge base to RDF/XML and RDFS/XML and the use of open source metadata browsers provide ready-made user interfaces that support both text- and facet-based search. This paper will present the Planetary Data System (PDS) data model as a use case and describe the import of the data model into an ontology tool

Licensed operating reactors

International Nuclear Information System (INIS)

1990-01-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors

International Nuclear Information System (INIS)

Hartfield, R.A.

1990-03-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors

International Nuclear Information System (INIS)

1989-08-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Criteria for selecting measures of plant information with application to nuclear reactors

International Nuclear Information System (INIS)

Kiguchi, T.; Sheridan, T.B.

1979-01-01

Criteria for selecting from alternative measures of plant information have been derived by information theory and the expected-value model of decision theory. Two criteria derived by information theory are minimum ''uncertainty'' of judgement of plant states, and maximum ''surprise'' of human operators. Two criteria derived by decision theory are maximum expected reward to operators, and maximum ratio of the maximum and next-to-maximum expected rewards. These four criteria have been evaluated experimentally in application to human monitoring problems of three ''plants'': an abstract laboratory exercise, a small research reactor, and a conventional nuclear power plant (high-fidelity training simulator). The results show that the criteria derived by decision theory are superior to the others from the viewpoint of correct rate of human subjects' judgement

Unification of Quantum and Gravity by Non Classical Information Entropy Space

Directory of Open Access Journals (Sweden)

Davide Fiscaletti

2013-09-01

Full Text Available A quantum entropy space is suggested as the fundamental arena describing the quantum effects. In the quantum regime the entropy is expressed as the superposition of many different Boltzmann entropies that span the space of the entropies before any measure. When a measure is performed the quantum entropy collapses to one component. A suggestive reading of the relational interpretation of quantum mechanics and of Bohm’s quantum potential in terms of the quantum entropy are provided. The space associated with the quantum entropy determines a distortion in the classical space of position, which appears as a Weyl-like gauge potential connected with Fisher information. This Weyl-like gauge potential produces a deformation of the moments which changes the classical action in such a way that Bohm’s quantum potential emerges as consequence of the non classical definition of entropy, in a non-Euclidean information space under the constraint of a minimum condition of Fisher information (Fisher Bohm- entropy. Finally, the possible quantum relativistic extensions of the theory and the connections with the problem of quantum gravity are investigated. The non classical thermodynamic approach to quantum phenomena changes the geometry of the particle phase space. In the light of the representation of gravity in ordinary phase space by torsion in the flat space (Teleparallel gravity, the change of geometry in the phase space introduces quantum phenomena in a natural way. This gives a new force to F. Shojai’s and A. Shojai’s theory where the geometry of space-time is highly coupled with a quantum potential whose origin is not the Schrödinger equation but the non classical entropy of a system of many particles that together change the geometry of the phase space of the positions (entanglement. In this way the non classical thermodynamic changes the classical geodetic as a consequence of the quantum phenomena and quantum and gravity are unified. Quantum

Thirteenth water reactor safety research information meeting: proceedings Volume 1

International Nuclear Information System (INIS)

Weiss, A.J.

1986-02-01

This six-volume report contains 151 papers out of the 178 that were presented at the Thirteenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 22-25, 1985. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-one different papers presented by researchers from Japan, Canada and eight European countries. The title of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume presents information on: risk analysis PRA application; severe accident sequence analysis; risk analysis/dependent failure analysis; and industry safety research

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Method and apparatus for monitoring the axial power distribution within the core of a nuclear reactor, exterior of the reactor

International Nuclear Information System (INIS)

Graham, K.F.; Gopal, R.

1978-01-01

A method and apparatus for establishing the axial flux distribution of a reactor core from monitored responses obtained exterior of the reactor is described. The monitored responses are obtained from at least three axially spaced flux responsive detectors that are positioned within proximity of the periphery of the reactor core. The detectors provide corresponding electrical outputs representative of the flux monitored. The axial height of the core is figuratively divided at a plurality of space coordinates sufficient to provide reconstruction in point representation of the relative flux shape along the core axis. The relative value of flux at each of the spaced coordinates is then established from a sum of the electrical outputs of the detectors, respectively, algebraically modified by a corresponding preestablished constant

Nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Prescott, R F; George, B V; Baglin, C J

1978-05-10

Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.; George, B.V.; Baglin, C.J.

1978-01-01

Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

Nuclear Reactors and Technology; (USA)

Energy Technology Data Exchange (ETDEWEB)

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Development of a nuclear information system for the MONJU Fast Breeder Reactor

International Nuclear Information System (INIS)

Metz, Peter

2000-01-01

At the MONJU Fast Breeder Reactor information is collected on a continuous basis. This information consists of measured data, design data, simulations data, maintenance data etc. which may be in any kind of electronic form, i.e. text documents, code input files, reports or even scanned documents. The amount and nature of these data has introduced the need for a software system, which will provide an efficient infrastructure for the maintenance of and operations on the data. Thus a Nuclear Information System for the MONJU Fast Breeder Reactor is under development. The system consists of remote databases hosting the information along with clients for handling them, remote clients providing the users with an interface and a local server for handling the client requests and the communication between the database and user clients. The system is composed of independent server, database and user modules, which communicates using the RMI-IIOP (Remote Method Invocation - Internet InterORB Protocol) technology. The RMI-IIOP is a CORBA (Common Object Request Broker Architecture) compliant subset of the RMI thereby facilitating the possibility of implementing the database and user modules in any kind of programming language and on any kind of operating system by providing a standard, platform independent communications interface. The user interface consists of dynamic HTML web pages which instantiates servlets in the user module when the user submits queries. The database module consists of controllers for handling the communication with the user module and database drivers for handling the connections with the databases. In this paper the overall system design and schemes for data flow and remote method invocations are presented and the requirements imposed on the system are discussed. (author)

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

International Nuclear Information System (INIS)

CJ Larkin; JD Edington; BJ Close

2006-01-01

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

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

Deep space telecommunications, navigation, and information management. Support of the space exploration initiative

Science.gov (United States)

Hall, Justin R.; Hastrup, Rolf C.

The United States Space Exploration Initiative (SEI) calls for the charting of a new and evolving manned course to the Moon, Mars, and beyond. This paper discusses key challenges in providing effective deep space telecommunications, navigation, and information management (TNIM) architectures and designs for Mars exploration support. The fundamental objectives are to provide the mission with means to monitor and control mission elements, acquire engineering, science, and navigation data, compute state vectors and navigate, and move these data efficiently and automatically between mission nodes for timely analysis and decision-making. Although these objectives do not depart, fundamentally, from those evolved over the past 30 years in supporting deep space robotic exploration, there are several new issues. This paper focuses on summarizing new requirements, identifying related issues and challenges, responding with concepts and strategies which are enabling, and, finally, describing candidate architectures, and driving technologies. The design challenges include the attainment of: 1) manageable interfaces in a large distributed system, 2) highly unattended operations for in-situ Mars telecommunications and navigation functions, 3) robust connectivity for manned and robotic links, 4) information management for efficient and reliable interchange of data between mission nodes, and 5) an adequate Mars-Earth data rate.

R-102, 1 Group Space-Independent Inverse Reactor Kinetics

International Nuclear Information System (INIS)

Kaganove, J.J.

1966-01-01

1 - Description of problem or function: Given the space-independent, one energy group reactor kinetics equations and the initial conditions, this program determines the time variation of reactivity required to produce the given input of flux-time data. 2 - Method of solution: Time derivatives of neutron density are obtained by application of (a) five-point quartic, (b) three-point parabolic, (c) five-point least-mean-square cubic, (d) five-point least-mean-square parabolic, or (e) five-point least-mean-square linear formulae to the neutron density or to the natural logarithm of the neutron density. Between each data point the neutron density is assumed to be (a) exponential*(third-order polynomial), (b) exponential, or (c) linear. Changes in reactivity between data points are obtained algebraically from the kinetics equations, neutron density derivatives, and the algebraic representation of neutron density. First and second time derivatives of the reactivity are obtained by use of any of the formulae applicable to the neutron density. 3 - Restrictions on the complexity of the problem: Maxima of - 50 delay groups; 1000 data points; 99 data blocks (A data block is a sequence of input points characterized by a fixed time-interval between points, a smoothing option, and a number of repetitions of the smoothing option)

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Kinetic parameters of the RB and RA reactors

Energy Technology Data Exchange (ETDEWEB)

Petrovic, M; Obradovic, D [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-12-15

In the paper the expressions for transfer functions of the zero power reactors, as well as power reactors of the RA reactor type are given, based on the space independent model. The modulation method for reactor transfer function measurements is explained. The results of the measurement and interpretation are given. The measurement were done on the RB and RA reactors in 'Boris Kidrich' Institute for Nuclear Sciences in Vincha (author)

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

Energy Technology Data Exchange (ETDEWEB)

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.

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

Effects of Weight Hourly Space Velocity and Catalyst Diameter on Performance of Hybrid Catalytic-Plasma Reactor for Biodiesel Synthesis over Sulphated Zinc Oxide Acid Catalyst

Directory of Open Access Journals (Sweden)

Luqman Buchori

2017-05-01

Full Text Available Biodiesel synthesis through transesterification of soybean oil with methanol on hybrid catalytic-plasma reactor over sulphated zinc oxide (SO42-/ZnO active acid catalyst was investigated. This research was aimed to study effects of Weight Hourly Space Velocity (WHSV and the catalyst diameter on performance of the hybrid catalytic-plasma reactor for biodiesel synthesis. The amount (20.2 g of active sulphated zinc oxide solid acid catalysts was loaded into discharge zone of the reactor. The WHSV and the catalyst diameter were varied between 0.89 to 1.55 min-1 and 3, 5, and 7 mm, respectively. The molar ratio of methanol to oil as reactants of 15:1 is fed to the reactor, while operating condition of the reactor was kept at reaction temperature of 65 oC and ambient pressure. The fatty acid methyl ester (FAME component in biodiesel product was identified by Gas Chromatography - Mass Spectrometry (GC-MS. The results showed that the FAME yield decreases with increasing WHSV. It was found that the optimum FAME yield was achieved of 56.91 % at WHSV of 0.89 min-1 and catalyst diameter of 5 mm and reaction time of 1.25 min. It can be concluded that the biodiesel synthesis using the hybrid catalytic-plasma reactor system exhibited promising the FAME yield. Copyright © 2017 BCREC Group. All rights reserved Received: 15th November 2016; Revised: 24th December 2016; Accepted: 16th February 2017 How to Cite: Buchori, L., Istadi, I., Purwanto, P. (2017. Effects of Weight Hourly Space Velocity and Catalyst Diameter on Performance of Hybrid Catalytic-Plasma Reactor for Biodiesel Synthesis over Sulphated Zinc Oxide Acid Catalyst. Bulletin of Chemical Reaction Engineering & Catalysis, 12 (2: 227-234 (doi:10.9767/bcrec.12.2.775.227-234 Permalink/DOI: http://dx.doi.org/10.9767/bcrec.12.2.775.227-234

Nuclear space power safety and facility guidelines study

International Nuclear Information System (INIS)

Mehlman, W.F.

1995-01-01

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

Exposure calculation code module for reactor core analysis: BURNER

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

Exposure calculation code module for reactor core analysis: BURNER

International Nuclear Information System (INIS)

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

Reactor container

International Nuclear Information System (INIS)

Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

1979-01-01

Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

Licensed operating reactors

International Nuclear Information System (INIS)

1990-01-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors

International Nuclear Information System (INIS)

1989-06-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units are provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear research reactors in the world. June 1988 ed.

International Nuclear Information System (INIS)

1988-01-01

This is the third edition of Reference Data Series No. 3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of June 1988, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the IAEA Research Reactor Data Base (RRDB) system. This system contains all the information and data previously published in the Agency's publication Power and Research Reactors in Member States as well as additional information. 12 figs, 19 tabs

Space Program Annual Report, For Approval

International Nuclear Information System (INIS)

TM Schaefer

2004-01-01

Knolls Atomic Power Laboratory (KAPL) (lead) has been requested by the Reference to create an unclassified report on the Prometheus Program's Jupiter Icy Moons Orbiter (JIMO) mission. This report is expected to be issued annually and be similar in level of content and scope to the NR Program's annual report ''The United States Naval Nuclear Propulsion Program'' (referred to as the Grey Book). The attachment to this letter provides a draft of the Prometheus Program report for NR review and approval. As stated in the Reference, a March 2005 issuance is planned following a coordinated NR Headquarter's review. The information contained in the attached report was obtained from open literature sources, NASA documents and Naval Reactors Program literature. The photographs contained in the report are drafts and their quality will be improved in the final version of the report. This report has been reviewed by the KAPL and Bettis Space Power Plant Staff and has been concurred with by the Manager of Space Power Plant (MJ Wollman) and the Manager of Bettis Reactor Engineering (C Eshelman)

Application of synthesis methods to two-dimensional fast reactor transient study

International Nuclear Information System (INIS)

Izutsu, Sadayuki; Hirakawa, Naohiro

1978-01-01

Space time synthesis and time synthesis codes were developed and applied to the space-dependent kinetics benchmark problem of a two-dimensional fast reactor model, and it was found both methods are accurate and economical for the fast reactor kinetics study. Comparison between the space time synthesis and the time synthesis was made. Also, in space time synthesis, the influence of the number of trial functions on the error and on the computing time and the effect of degeneration of expansion coefficients are investigated. The matrix factorization method is applied to the inversion of the matrix equation derived from the synthesis equation, and it is indicated that by the use of this scheme space-dependent kinetics problem of a fast reactor can be solved efficiently by space time synthesis. (auth.)

Nuclear-electric power in space

International Nuclear Information System (INIS)

Truscello, V.C.; Davis, H.S.

1984-01-01

Because direct-broadcast satellites, air-traffic-control radar satellites, industrial processing on subsequent versions of the space station, and long range excursions to other planets using nuclear-electric propulsion systems, all space missions for which current power-supply systems are not sufficient. NASA and the DOE therefore have formed a joint program to develop the technology required for nuclear-reactor space power plants. After investigating potential space missions in the given range, the project will develop the technology to build such systems. High temperatures pose problems, ''hot shoes'' and ''cold shoes'', a Stirling engine dynamic system, and critical heat-transfer problems are all discussed. The nuclear reactor system for space as now envisioned is schematicized

Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

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

Science.gov (United States)

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.

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

A Reference Architecture for Space Information Management

Science.gov (United States)

Mattmann, Chris A.; Crichton, Daniel J.; Hughes, J. Steven; Ramirez, Paul M.; Berrios, Daniel C.

2006-01-01

We describe a reference architecture for space information management systems that elegantly overcomes the rigid design of common information systems in many domains. The reference architecture consists of a set of flexible, reusable, independent models and software components that function in unison, but remain separately managed entities. The main guiding principle of the reference architecture is to separate the various models of information (e.g., data, metadata, etc.) from implemented system code, allowing each to evolve independently. System modularity, systems interoperability, and dynamic evolution of information system components are the primary benefits of the design of the architecture. The architecture requires the use of information models that are substantially more advanced than those used by the vast majority of information systems. These models are more expressive and can be more easily modularized, distributed and maintained than simpler models e.g., configuration files and data dictionaries. Our current work focuses on formalizing the architecture within a CCSDS Green Book and evaluating the architecture within the context of the C3I initiative.

A Concept of Constructing a Common Information Space for High Tech Programs Using Information Analytical Systems

Science.gov (United States)

Zakharova, Alexandra A.; Kolegova, Olga A.; Nekrasova, Maria E.

2016-04-01

The paper deals with the issues in program management used for engineering innovative products. The existing project management tools were analyzed. The aim is to develop a decision support system that takes into account the features of program management used for high-tech products: research intensity, a high level of technical risks, unpredictable results due to the impact of various external factors, availability of several implementing agencies. The need for involving experts and using intelligent techniques for information processing is demonstrated. A conceptual model of common information space to support communication between members of the collaboration on high-tech programs has been developed. The structure and objectives of the information analysis system “Geokhod” were formulated with the purpose to implement the conceptual model of common information space in the program “Development and production of new class mining equipment - “Geokhod”.

Reactor power system deployment and startup

International Nuclear Information System (INIS)

Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

International Nuclear Information System (INIS)

Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

2016-01-01

Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

Reactor technology. Progress report, January--March 1978

International Nuclear Information System (INIS)

Warren, J.L.

1978-07-01

Progress is reported in eight program areas. The nuclear Space Electric Power Supply Program examined safety questions in the aftermath of the COSMOS 954 incident, examined the use of thermoelectric converters, examined the neutronic effectiveness of various reflecting materials, examined ways of connecting heat pipes to one another, studied the consequences of the failure of one heat pipe in the reactor core, and did conceptual design work on heat radiators for various power supplies. The Heat Pipe Program reported progress in the design of ceramic heat pipes, new application of heat pipes to solar collectors, and final performance tests of two pipes for HEDL applications. Under the Nuclear Process Heat Program, work continues on computer codes to model a pebble bed high-temperature gas-cooled reactor, adaptation of a set of German reactor calculation codes to use on U.S. computers, and a parametric study of a certain resonance integral required in reactor studies. Under the Nonproliferation Alternative Sources Assessment Program LASL has undertaken an evaluation of a study of gaseous core reactors by Southern Science Applications, Inc. Independently LASL has developed a proposal for a comprehensive study of gaseous uranium-fueled reactor technology. The Plasma Core Reactor Program has concentrated on restacking the beryllium reflector and redesigning the nuclear control system. The status of and experiments on four critical assemblies, SKUA, Godiva IV, Big Ten, and Flattop, are reported. The Nuclear Criticality Safety Program carried out several tasks including conducting a course, doing several annual safety reviews and evaluating the safety of two Nevada test devices. During the quarter one of the groups involved in reactor technology has acquired responsibility for the operation of a Cockroft-Walton accelerator. The present report contains information on the use of machine and improvements being made in its operation

Some scoping experiments for a space reactor

International Nuclear Information System (INIS)

Alexander, C.A.; Ogden, J.S.

1983-01-01

Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO 2 and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor

An advanced method of heterogeneous reactor theory

International Nuclear Information System (INIS)

Kochurov, B.P.

1994-08-01

Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

New experimental space for irradiating samples by RA reactor fast neutron flux at temperatures up to 100 deg C

International Nuclear Information System (INIS)

Pavicevic, M.; Novakovic, M.; Zecevic, V.

1961-01-01

The objective of this paper is to present adaptation of the RA reactor which would enable samples irradiation by fast neutrons and describe new experimental possibilities. New experimental space was achieved using hollow fuel elements which have been reconstructed to enable placement of irradiation capsules inside the tube. This paper includes thermal analysis and describes problems related to operation, safety and radiation protection issues which arise from using reconstructed fuel elements

Space nuclear power and man's extraterrestrial civilization

International Nuclear Information System (INIS)

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

Method and apparatus for removing radioactive gases from a nuclear reactor

International Nuclear Information System (INIS)

Frumerman, R.; Brown, W.W.

1975-01-01

A description is given of a method for removing radioactive gases from a nuclear reactor including the steps of draining coolant from a nuclear reactor to a level just below the coolant inlet and outlet nozzles to form a vapor space and then charging the space with an inert gas, circulating coolant through the reactor to assist the release of radioactive gases from the coolant into the vapor space, withdrawing the radioactive gases from the vapor space by a vacuum pump which then condenses and separates water from gases carried forward by the vacuum pump, discharging the water to a storage tank and supplying the separated gases to a gas compressor which pumps the gases to gas decay tanks. After the gases in the decay tanks lose their radioactive characteristics, the gases may be discharged to the atmosphere or returned to the reactor for further use

Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2015-05-15

Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

Background information document to support NESHAPS rulemaking on nuclear power reactors. Draft report

International Nuclear Information System (INIS)

Colli, A.; Conklin, C.; Hoffmeyer, D.

1991-08-01

The purpose of this Background Information Document (BID) is to present information relevant to the Administrator of the Environmental Protection Agency's (EPA) reconsideration of the need for a NESHAP to control radionuclides emitted to the air from commercial nuclear power reactors. The BID presents information on the relevant portions of the regulatory framework that NRC has implemented for nuclear power plant licensees, under the authority of the Atomic Energy Act, as amended, to protect the public's health and safety. To provide context, it summarizes the rulemaking history for Subpart I. It then describes NRC's regulatory program for routine atmospheric emissions of radionuclides and evaluates the doses caused by actual airborne emissions from nuclear power plants, including releases resulting from anticipated operational occurrences

Nuclear power reactors in the world. Apr 1985 ed.

International Nuclear Information System (INIS)

1985-01-01

This is the fifth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which replaces the Agency's publication Power Reactors in Member States. This bulletin contains the following summarized information on nuclear power reactors in the world: General information as of the end of 1984 on reactors operating or under construction and such additional information on planned and shutdown reactors as is available; Performance data on major reactor types operating in the Agency's Member States. The information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of and operating experience with reactors

Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1993-03-01

This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan

Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

International Nuclear Information System (INIS)

Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

2008-01-01

The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

A fast spectrum dual path flow cermet reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

Science.gov (United States)

El-Genk, Mohamed S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

International Nuclear Information System (INIS)

El-genk, M.S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

Nuclear reactor built, being built, or planned

International Nuclear Information System (INIS)

1991-06-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1990. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly

Information Presentation: Human Research Program - Space Human Factors and Habitability, Space Human Factors Engineering Project

Science.gov (United States)

Holden, Kristina L.; Sandor, Aniko; Thompson, Shelby G.; Kaiser, Mary K.; McCann, Robert S.; Begault, D. R.; Adelstein, B. D.; Beutter, B. R.; Wenzel, E. M.; Godfroy, M.;

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

International Nuclear Information System (INIS)

Haloulakos, V.E.

1991-01-01

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

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

Science.gov (United States)

Haloulakos, V. E.

1991-01-01

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

Space Geodesy Project Information and Configuration Management Procedure

Science.gov (United States)

Merkowitz, Stephen M.

2016-01-01

This plan defines the Space Geodesy Project (SGP) policies, procedures, and requirements for Information and Configuration Management (CM). This procedure describes a process that is intended to ensure that all proposed and approved technical and programmatic baselines and changes to the SGP hardware, software, support systems, and equipment are documented.

The complete information for phenomenal distributed parameter control of multicomponent chemical processes in gas, fluid and solid phase

International Nuclear Information System (INIS)

Niemiec, W.

1985-01-01

A constitutive mathematical model of distributed parameters of multicomponent chemical processes in gas, fluid and solid phase is utilized to the realization of phenomenal distributed parameter control of these processes. Original systems of partial differential constitutive state equations, in the following derivative forms /I/, /II/ and /III/ are solved in this paper from the point of view of information for phenomenal distributed parameter control of considered processes. Obtained in this way for multicomponent chemical processes in gas, fluid and solid phase: -dynamical working space-time characteristics/analytical solutions in working space-time of chemical reactors/, -dynamical phenomenal Green functions as working space-time transfer functions, -statical working space characteristics /analytical solutions in working space of chemical reactors/, -statical phenomenal Green functions as working space transfer functions, are applied, as information for realization of constitutive distributed parameter control of mass, energy and momentum aspects of above processes. Two cases are considered by existence of: A/sup o/ - initial conditions, B/sup o/ - initial and boundary conditions, for multicomponent chemical processes in gas, fluid and solid phase

BWR type reactors

International Nuclear Information System (INIS)

Hayashi, Katsuhisa; Watanabe, Shigeru.

1983-01-01

Purpose: To simplify the structure of control rod driving systems, as well as improve the safety and maintainability thereof. Constitution: Control-rod-guide tubes are disposed vertically above the reactor core and control-rod drives are disposed further thereabove, by which the control rods are moved upwardly and downwardly from above the reactor core through the guide tubes. Further, a partitioning cylinder is provided between the inner cirumferential wall at the upper portion of a pressure vessel and the control-rod-guide tubes and a gas-liquid separator is disposed to the space between the partitioning cylinder and the pressure vessel wall, to which steams generated in the reactor core are introduced. In such a structure of the reactor, since all of the control rods are inserted or extracted by the control rod drive system from above the reactor core, if the control rod drives or the likes should fail and accidentally drop the control rods, they exert in the direction of suppressing the nuclear reaction, whereby the safety can be improved. (Sekiya, K.)

Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

International Nuclear Information System (INIS)

Atkins, D.F.; Lillie, A.F.

1983-01-01

As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

Reactor-specific spent fuel discharge projections: 1985 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

Space in Numerical and Ordinal Information: A Common Construct?

Directory of Open Access Journals (Sweden)

Philipp Alexander Schroeder

2017-12-01

Full Text Available Space is markedly involved in numerical processing, both explicitly in instrumental learning and implicitly in mental operations on numbers. Besides action decisions, action generations, and attention, the response-related effect of numerical magnitude or ordinality on space is well documented in the Spatial-Numerical Associations of Response Codes (SNARC effect. Here, right- over left-hand responses become relatively faster with increasing magnitude positions. However, SNARC-like behavioral signatures in non-numerical tasks with ordinal information were also observed and inspired new models integrating seemingly spatial effects of ordinal and numerical metrics. To examine this issue further, we report a comparison between numerical SNARC and ordinal SNARC-like effects to investigate group-level characteristics and individual-level deductions from generalized views, i.e., convergent validity. Participants solved order-relevant (before/after classification and order-irrelevant tasks (font color classification with numerical stimuli 1-5, comprising both magnitude and order information, and with weekday stimuli, comprising only ordinal information. A small correlation between magnitude- and order-related SNARCs was observed, but effects are not pronounced in order-irrelevant color judgments. On the group level, order-relevant spatial-numerical associations were best accounted for by a linear magnitude predictor, whereas the SNARC effect for weekdays was categorical. Limited by the representativeness of these tasks and analyses, results are inconsistent with a single amodal cognitive mechanism that activates space in mental processing of cardinal and ordinal information alike. A possible resolution to maintain a generalized view is proposed by discriminating different spatial activations, possibly mediated by visuospatial and verbal working memory, and by relating results to findings from embodied numerical cognition.