WorldWideScience

Sample records for space fission reactor

  1. 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.

  2. 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

  3. 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

  4. HAC and fission reactors

    International Nuclear Information System (INIS)

    Fujiwara, I.; Moriyama, H.; Tachikawa, E.

    1984-01-01

    In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

  5. 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)

  6. Space Fission System Test Effectiveness

    International Nuclear Information System (INIS)

    Houts, Mike; Schmidt, Glen L.; Van Dyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky; Salvail, Pat; Harper, Roger

    2004-01-01

    Space fission technology has the potential to enable rapid access to any point in the solar system. If fission propulsion systems are to be developed to their full potential, however, near-term customers need to be identified and initial fission systems successfully developed, launched, and utilized. One key to successful utilization is to develop reactor designs that are highly testable. Testable reactor designs have a much higher probability of being successfully converted from paper concepts to working space hardware than do designs which are difficult or impossible to realistically test. ''Test Effectiveness'' is one measure of the ability to realistically test a space reactor system. The objective of this paper is to discuss test effectiveness as applied to the design, development, flight qualification, and acceptance testing of space fission systems. The ability to perform highly effective testing would be particularly important to the success of any near-term mission, such as NASA's Jupiter Icy Moons Orbiter, the first mission under study within NASA's Project Prometheus, the Nuclear Systems Program

  7. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  8. Fission reactor container

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1991-01-01

    Cooling water is sent without using dynamic equipments upon loss of coolants accident in a pressure vessel by improving an arrangement of a nuclear reactor pressure vessel. That is, a containing space is formed at the center of a suppression chamber for storing cooling water while being partitioned with each other, in which the pressure vessel is placed. Further, a water reservoir is formed above the pressure vessel. Then a water discharge pipe is connected to the reservoir for submerging the stored water over the pressure vessel upon occurrence of loss of coolants accident. Further, a water injection pipe is disposed between the pressure suppression chamber and the pressure vessel for injecting the cooling water in the pressure suppression chamber to the reactor core of the pressure vessel by the difference of a water head upon loss of coolants accident. With such a constitution, the pressure vessel has high earthquake proofness. Further, upon loss of coolants accident of the pressure vessel, the cooling water in the reservoir is discharged to submerge and cool the pressure vessel efficiently. Further, the reactor core of the pressure vessel can certainly be cooled by the cooling water of the pressure suppression chamber without relying on dynamic equipments. (I.S.)

  9. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  10. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  11. Fission reactor control rod

    International Nuclear Information System (INIS)

    Irie, Tomoo.

    1991-01-01

    The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

  12. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  13. Natural fission reactors - the Oklo phenomenon

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Overview describes the discovery of the site, location of the reactors and site geology and discusses the permanence of fission products, nuclear reaction control mechanisms and trace concentrations of elements that act as poisons. (Author)

  14. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  15. Reference reactor module for NASA's lunar surface fission power system

    International Nuclear Information System (INIS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  16. Transmutation of Tc-99 in fission reactors

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1994-12-01

    Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

  17. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  18. Mirror hybrid (fusion--fission) reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Neef, W.S.; Devoto, R.S.; Galloway, T.R.; Fink, J.H.; Schultz, K.R.; Culver, D.; Rao, S.

    1977-10-01

    The reference mirror hybrid reactor design performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. The resulting reactor may thus be viewed as a comparatively near-term goal of the fusion program, and we project improved performance for the hybrid in the future as more advanced technology becomes available

  19. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  20. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  1. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Marwick, E.F.

    1978-01-01

    Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

  2. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.; Masson, M.; Briec, M.

    1986-09-01

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

  3. Reactor physics and thermodynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1990-01-01

    Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  4. Utilization of fission reactors for fusion engineering testing

    International Nuclear Information System (INIS)

    Deis, G.A.; Miller, L.G.

    1985-01-01

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

  5. Nuclear materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.; Schumacher, G.

    1992-01-01

    This volume brings together 47 papers from scientists involved in the fabrication of new nuclear fuels, in basic research of nuclear materials, their application and technology as well as in computer codes and modelling of fuel behaviour. The main emphasis is on progress in the development of non -oxide fuels besides reporting advances in the more conventional oxide fuels. The two currently performed large reactor safety programmes CORA and PHEBUS-FP are described in invited lectures. The contributions review basic property measurements, as well as the present state of fuel performance modelling. The performance of today's nuclear fuel, hence UO 2 , at high burnup is also reviewed with particular emphasis on the recently observed phenomenon of grain subdivision in the cold part of the oxide fuel at high burnup, the so-called 'rim' effect. Similar phenomena can be simulated by ion implantation in order to better elucidate the underlying mechanism and reviews on high resolution electron microscopy provide further information. The papers will provide a useful treatise of views, ideas and new results for all those scientists and engineers involved in the specific questions of current nuclear waste management

  6. Search for other natural fission reactors

    International Nuclear Information System (INIS)

    Apt, K.E.; Balagna, J.P.; Bryant, E.A.; Cowan, G.A.; Daniels, W.R.; Vidale, R.J.

    1977-01-01

    Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m 3 and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient 235 U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor

  7. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-09-01

    New options for the development of the nuclear energy economy which might become available by a successful development of fusion-breeders or fusion-fission hybrid power reactors, identified and their nonproliferative attributes are discussed. The more promising proliferation-resistance ettributes identified include: (1) Justification for a significant delay in the initiation of fuel processing, (2) Denaturing the plutonium with 238 Pu before its use in power reactors of any kind, and (3) Making practical the development of denatured uranium fuel cycles and, in particular, denaturing the uranium with 232 U. Fuel resource utilization, time-table and economic considerations associated with the use of fusion-breeders are also discussed. It is concluded that hybrid reactors may enable developing a nuclear energy economy which is more proliferation resistant than possible otherwise, whileat the same time, assuring high utilization of t he uranium and thorium resources in an economically acceptable way. (author)

  8. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  9. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  10. Hybrid fission-fusion nuclear reactors

    International Nuclear Information System (INIS)

    Zucchetti, Massimo

    2011-01-01

    A fusion-fission hybrid could contribute to all components of nuclear power - fuel supply, electricity production, and waste management. The idea of the fusion-fission hybrid is many decades old. Several ideas, both new and revisited, have been investigated by hybrid proponents. These ideas appear to have attractive features, but they require various levels of advances in plasma science and fusion and nuclear technology. As a first step towards the development of hybrid reactors, fusion neutron sources can be considered as an option. Compact high-field tokamaks can be a candidate for being the neutron source in a fission-fusion hybrid, essentially due to their design characteristics, such as compact dimensions, high magnetic field, flexibility of operation. This study presents the development of a tokamak neutron source for a material testing facility using an Ignitor-based concept. The computed values show the potential of this neutron-rich device for fusion materials testing. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa. (Author)

  11. Thermal Energetic Reactor with High Reproduction of Fission Materials

    Directory of Open Access Journals (Sweden)

    Vladimir M. Kotov

    2012-01-01

    On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  12. Legal and Regulatroy Obstacles to Nuclear Fission Technology in Space

    Science.gov (United States)

    Force, Melissa K.

    2013-09-01

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

  13. Matching of dense plasma focus devices with fission reactors

    International Nuclear Information System (INIS)

    Harms, A.A.; Heindler, M.

    1978-01-01

    The potential role of dense plasma focus devices as compact neutron sources for fissile fuel breeding in conjunction with existing fission reactors is considered. It is found that advanced plasma focus devices can be used effectively in conjunction with neutronically efficient fission reactors to constitute ''self-sufficient'' breeders. Correlations among the various parameters such as the power output and conversion ratio of the fission reactor with the neutron yield and capacitor bank energy of the dense plasma focus device are presented and discussed

  14. Innovative fission reactors for this century

    International Nuclear Information System (INIS)

    Minguez, E.

    2007-01-01

    of the 21st Century both innovative fission reactors and fusion reactors. For 2025, it seems that many countries of EU will have to construct NPPs until 40 GWe: France, UK, Germany, North Europe, Russia, Spain, Rumania and Turkey, between others. The viability of these innovative concepts will be presented in this paper

  15. Enabling the Use of Space Fission Propulsion Systems

    International Nuclear Information System (INIS)

    Mike Houts; Melissa Van Dyke; Tom Godfroy; James Martin; Kevin Pedersen; Ricky Dickens; Ivana Hrbud; Leo Bitteker; Bruce Patton; Suman Chakrabarti; Joe Bonometti

    2000-01-01

    This paper gives brief descriptions of advantages of fission technology for reaching any point in the solar system and of earlier efforts to develop space fission propulsion systems, and gives a more detailed description of the safe, affordable fission engine (SAFE) concept being pursued at the National Aeronautics and Space Administration's Marshall Space Flight Center

  16. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  17. The LOFA analysis of fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Z.-C.; Xie, H.

    2014-01-01

    The fusion-fission hybrid energy reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc, with the fusion neutron source striking the subcritical blanket. The passive safety system, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (author)

  18. Nuclear data in the problem of fission reactor decommissioning

    International Nuclear Information System (INIS)

    Manokhin, V.N.; Kulagin, N.T.

    1993-01-01

    This report presents a review of the works published in Russia during last several years and devoted to the problem of nuclear data and calculations of nuclear facilities activation for fission reactor decommissioning. 6 refs

  19. Options for development of space fission propulsion systems

    International Nuclear Information System (INIS)

    Houts, Mike; Van Dyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana

    2001-01-01

    Fission technology can enable rapid, affordable access to any point in the solar system. Potential fission-based transportation options include high specific power continuous impulse propulsion systems and bimodal nuclear thermal rockets. Despite their tremendous potential for enhancing or enabling deep space and planetary missions, to date space fission systems have only been used in Earth orbit. The first step towards utilizing advanced fission propulsion systems is development of a safe, near-term, affordable fission system that can enhance or enable near-term missions of interest. An evolutionary approach for developing space fission propulsion systems is proposed

  20. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    Juhl, N.H.; Marwick, E.F.

    1983-01-01

    A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

  1. Thermal Energetic Reactor with High Reproduction of Fission Materials

    International Nuclear Information System (INIS)

    Kotov, V.M.

    2012-01-01

    Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  2. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  3. Chemistry of fission product iodine under nuclear reactor accident conditions

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

  4. On the safety of conceptual fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

    1979-01-01

    A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

  5. Fission power: a search for a ''second-generation'' reactor

    International Nuclear Information System (INIS)

    Hovingh, J.

    1985-01-01

    This report touches on the history of US fission reactors and explores the current technical status of such reactors around the world, including experimental reactors. Its purpose is to identify, evaluate, and rank the most promising concepts among existing reactors, proposed but unadopted designs, and what can be described as ''new'' concepts. Also discussed are such related concerns as utility requirements and design considerations. The report concludes with some recommendations for possible future LLNL involvement

  6. Reference reactor module for NASA's lunar surface fission power system

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  7. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

    The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

  8. Brief review of the fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1977-01-01

    Much of the conceptual framework of present day fusion-fission hybrid reactors is found in the original work of the early 1950's. Present day motivations for development are quite different. The role of the hybrid reactor is discussed as well as the current activities in the development program

  9. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

  10. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  11. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  12. On fusion and fission breeder reactors

    International Nuclear Information System (INIS)

    Brandt, B.; Schuurman, W.; Klippel, H.Th.

    1981-02-01

    Fast breeder reactors and fusion reactors are suitable candidates for centralized, long-term energy production, their fuel reserves being practically unlimited. The technology of a durable and economical fusion reactor is still to be developed. Such a development parallel with the fast breeder is valuable by reasons of safety, proliferation, new fuel reserves, and by the very broad potential of the development of the fusion reactor. In order to facilitate a discussion of these aspects, the fusion reactor and the fast breeder reactor were compared in the IIASA-report. Aspects of both reactor systems are compared

  13. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-09-01

    A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

  14. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements

  15. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  16. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  17. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  18. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  19. Updated comparison of economics of fusion reactors with advanced fission reactors

    International Nuclear Information System (INIS)

    Delene, J.G.

    1990-01-01

    The projected cost of electricity (COE) for fusion is compared with that from current and advanced nuclear fission and coal-fired plants. Fusion cost models were adjusted for consistency with advanced fission plants and the calculational methodology and cost factors follow guidelines recommended for cost comparisons of advanced fission reactors. The results show COEs of about 59--74 mills/kWh for the fusion designs considered. In comparison, COEs for future fission reactors are estimated to be in the 43--54 mills/kWh range with coal-fired plant COEs of about 53--69 mills/kWh ($2--3/GJ coal). The principal cost driver for the fusion plants relative to fission plants is the fusion island cost. Although the estimated COEs for fusion are greater than those for fission or coal, the costs are not so high as to preclude fusion's competitiveness as a safe and environmentally sound alternative

  20. Thermochemical data for reactor materials and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1990-01-01

    This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

  1. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  2. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  3. Safety device for nuclear fission reactors

    International Nuclear Information System (INIS)

    Brownlee, M.L.

    1982-01-01

    A plurality of radially arranged and neutron absorbing baffles are stacked in vertical sets under the fuel core assemblies, and the whole enclosed in a bottle shaped containment vessel. The radially arranged baffles of each set extend vertically, and each set has double the number of baffles as the set above it in the stack. A melt-down of a fuel core assembly drops the fissioning nuclear fuel into the stacked sets of baffles, there, as it passes through, to be progressively divided, redivided and dispersed in smaller and smaller masses between the doubling number of baffles in safe fuel pellet size. Neutron absorbing containment prevents contamination of the environment and together with cooling means stops fissioning of fuel

  4. Neutron and thermal dynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1989-01-01

    In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  5. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a 233 U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

  6. Space nuclear reactor safety

    International Nuclear Information System (INIS)

    Damon, D.; Temme, M.; Brown, N.

    1990-01-01

    Definition of safety requirements and design features of the SP-100 space reactor power system has been guided by a mission risk analysis. The analysis quantifies risk from accidental radiological consequences for a reference mission. Results show that the radiological risk from a space reactor can be made very low. The total mission risk from radiological consequences for a shuttle-launched, earth orbit SP-100 mission is estimated to be 0.05 Person-REM (expected values) based on a 1 mREM/yr de Minimus dose. Results are given for each mission phase. The safety benefits of specific design features are evaluated through risk sensitivity analyses

  7. Conceptual Analysis of Fission Fragment Magnetic Collimator Reactors

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.; Parish, Theodore A.

    2002-01-01

    As part of the current research work within the US DOE NERI Direct Electricity Conversion (DEC) Project on methods for utilizing direct electricity conversion in nuclear reactors, a detailed study of a Fission Fragment Magnetic Collimator Reactor (FFMCR) has been performed. The FFMCR concept is an advanced DEC system that combines advantageous design solutions proposed for application in both fission and fusion reactors. The present study was focused on determining the electrical efficiency and other important operational aspects of the FFMCR concept. In principle, acceptable characteristics have been demonstrated, and results obtained are presented in the paper. Technological visibility of the FFMCR concept and required further design development are discussed. Preliminary characteristics of the promising design are outlined. (authors)

  8. Developments and Tendencies in Fission Reactor Concepts

    Science.gov (United States)

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

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  9. Status on potential of advanced fission reactors

    International Nuclear Information System (INIS)

    L-Zaleski, C.P.

    1978-01-01

    In this short lecture, only two types of reactors will be discussed: the liquid metal fast breeder reactors (LMFBR) and the high temperature reactors (HTR). This does not mean that other very interesting concepts do not exist, but there are or proven light water reactors and heavy water reactors or has not reached the state of industrial development like molten-salt or gas breeder reactors. In discussing any types of industrial development, it seems to me useful, first to indicate the reasons or motivations for this development. Then I will give a short historical review and analysis of what has been done up to now. For HTR's a very brief status report will be presented. For LMFBR's, I will give indications of experience gained with demonstration plants and more specifically with Phenix, before listing the most important technical problems which still need more work to be fully solved. Finally, I will briefly discuss the economic status and perspectives of LMFBR's and will mention the public acceptance problem

  10. Fission product release from TRIGA-LEU reactor fuels

    International Nuclear Information System (INIS)

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-01-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  11. Fission reactor recycling pump handling device

    International Nuclear Information System (INIS)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru.

    1991-01-01

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.)

  12. Fission reactor recycling pump handling device

    Energy Technology Data Exchange (ETDEWEB)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru

    1991-06-24

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.).

  13. Overview of standards subcommittee 8, fissionable materials outside reactors

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1996-01-01

    The American Nuclear Society's Standards Subcommittee 8, titled open-quotes Fissionable Materials Outside Reactors,close quotes has worked for the past 35 yr to prepare and promote standards on nuclear criticality safety for the handling, processing, storing, and transportation of fissionable materials outside reactors. The reader is referred to the Transactions of the American Nuclear Society, Vols. 39 (1981) and 64 (1991), for previous papers associated with ANS-8 poster sessions. In addition to discussions on the then-current standards, the reader will find articles on working group efforts that never materialized into standards, such as proposed 8.13, open-quotes Use of the Solid-Angle Method in Nuclear Criticality Safety,close quotes and on applications and critiques of current standards. The paper by McLendon in Vol. 39 is particularly interesting as an overview of the early history of ANS-8 and its standards

  14. Expected value of finite fission chain lengths of pulse reactors

    International Nuclear Information System (INIS)

    Liu Jianjun; Zhou Zhigao; Zhang Ben'ai

    2007-01-01

    The average neutron population necessary for sponsoring a persistent fission chain in a multiplying system, is discussed. In the point reactor model, the probability function θ(n, t 0 , t) of a source neutron at time t 0 leading to n neutrons at time t is dealt with. The non-linear partial differential equation for the probability generating function G(z; t 0 , t) is derived. By solving the equation, we have obtained an approximate analytic solution for a slightly prompt supercritical system. For the pulse reactor Godiva-II, the mean value of finite fission chain lengths is estimated in this work and shows that the estimated value is reasonable for the experimental analysis. (authors)

  15. Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

  16. 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

  17. Role of fission gas release in reactor licensing

    International Nuclear Information System (INIS)

    1975-11-01

    The release of fission gases from oxide pellets to the fuel rod internal voidage (gap) is reviewed with regard to the required safety analysis in reactor licensing. Significant analyzed effects are described, prominent gas release models are reviewed, and various methods used in the licensing process are summarized. The report thus serves as a guide to a large body of literature including company reports and government documents. A discussion of the state of the art of gas release analysis is presented

  18. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  19. Benchmark experiments at ASTRA facility on definition of space distribution of 235U fission reaction rate

    International Nuclear Information System (INIS)

    Bobrov, A. A.; Boyarinov, V. F.; Glushkov, A. E.; Glushkov, E. S.; Kompaniets, G. V.; Moroz, N. P.; Nevinitsa, V. A.; Nosov, V. I.; Smirnov, O. N.; Fomichenko, P. A.; Zimin, A. A.

    2012-01-01

    Results of critical experiments performed at five ASTRA facility configurations modeling the high-temperature helium-cooled graphite-moderated reactors are presented. Results of experiments on definition of space distribution of 235 U fission reaction rate performed at four from these five configurations are presented more detail. Analysis of available information showed that all experiments on criticality at these five configurations are acceptable for use them as critical benchmark experiments. All experiments on definition of space distribution of 235 U fission reaction rate are acceptable for use them as physical benchmark experiments. (authors)

  20. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.

    1995-01-01

    Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

  1. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

  2. Structural materials issues for the next generation fission reactors

    Science.gov (United States)

    Chant, I.; Murty, K. L.

    2010-09-01

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

  3. Fusion--fission hybrid reactors based on the laser solenoid

    International Nuclear Information System (INIS)

    Steinhauer, L.C.; Taussig, R.T.; Quimby, D.C.

    1976-01-01

    Fusion-fission reactors, based on the laser solenoid concept, can be much smaller in scale than their pure fusion counterparts, with moderate first-wall loading and rapid breeding capabilities (1 to 3 tonnes/yr), and can be designed successfully on the basis of classical plasma transport properties and free-streaming end-loss. Preliminary design information is presented for such systems, including the first wall, pulse coil, blanket, superconductors, laser optics, and power supplies, accounting for the desired reactor performance and other physics and engineering constraints. Self-consistent point designs for first and second generation reactors are discussed which illustrate the reactor size, performance, component parameters, and the level of technological development required

  4. Thermodynamic cycle calculations for a pumped gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.

    1991-01-01

    Finite and 'infinitesimal' thermodynamic cycle calculations have been performed for a 'solid piston' model of a pumped Gaseous Core Fission Reactor with dissociating reactor gas, consisting of Uranium, Carbon and Fluorine ('UCF'). In the finite cycle calculations the influence has been investigated of several parameters on the thermodynamics of the system, especially on the attainable direct (nuclear to electrical) energy conversion efficiency. In order to facilitate the investigation of the influence of dissociation, a model gas, 'Modelium', was developed, which approximates, in a simplified, analytical way, the dissociation behaviour of the 'real' reactor gas. Comparison of the finite cycle calculation results with those of a so-called infinitesimal Otto cycle calculation leads to the conclusion that the conversion efficiency of a finite cycle can be predicted, without actually performing the finite cycle calculation, with reasonable accuracy, from the so-called 'infinitesimal efficiency factor', which is determined only by the thermodynamic properties of the reactor gas used. (author)

  5. Fission product poisoning in KS-150 reactor operation

    International Nuclear Information System (INIS)

    Rana, S.B.

    1978-01-01

    A three-dimensional model of the KS-150 reactor was used to study reactivity changes induced by reactor poisoning with fission products Xe 135 and Sm 149 . A comparison of transients caused by the poisoning showed the following differences: (1) the duration of the transient Xe poisoning (2 days) is shorter by one order of magnitude than the duration of Sm poisoning (20 days); however, the level of Xe poisoning is greater approximately by one order than the level of the Sm poisoning; (2) the level of steady-state Xe poisoning depends on the output level of the reactor; steady-state Sm poisoning does not depend on this level; (3) following reactor shutdown Xe poisoning may increase to the maximum value of up to Δrhosub(Xe)=20% and will then gradually decrease; Sm poisoning may reach maximum values of up to Δrhosub(Sm)=2% and does not decrease. (J.B.)

  6. Nuclear data requirements for fission reactor decommissioning

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1993-01-01

    The meeting was attended by 13 participants from 8 Member States and 2 International Organizations who reviewed the status of the nuclear data libraries and computer codes used to calculate the radioactive inventory in the reactor unit components for the decommissioning purposes. Nuclides and nuclear reactions important for determination of the radiation fields during decommissioning and for the final disposal of radioactive waste from the decommissioned units were identified. Accuracy requirements for the relevant nuclear data were considered. The present publication contains the text of the reports by the participants and their recommendations to the Nuclear Data Section of the IAEA. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  7. Neutron dosimetry for radiation damage in fission and fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1979-01-01

    The properties of materials subjected to the intense neutron radiation fields characteristic of fission power reactors or proposed fusion energy devices is a field of extensive current research. These investigations seek important information relevant to the safety and economics of nuclear energy. In high-level radiation environments, neutron metrology is accomplished predominantly with passive techniques which require detailed knowledge about many nuclear reactions. The quality of neutron dosimetry has increased noticeably during the past decade owing to the availability of new data and evaluations for both integral and differential cross sections, better quantitative understanding of radioactive decay processes, improvements in radiation detection technology, and the development of reliable spectrum unfolding procedures. However, there are problems caused by the persistence of serious integral-differential discrepancies for several important reactions. There is a need to further develop the data base for exothermic and low-threshold reactions needed in thermal and fast-fission dosimetry, and for high-threshold reactions needed in fusion-energy dosimetry. The unsatisfied data requirements for fission reactor dosimetry appear to be relatively modest and well defined, while the needs for fusion are extensive and less well defined because of the immature state of fusion technology. These various data requirements are examined with the goal of providing suggestions for continued dosimetry-related nuclear data research

  8. Hefei experimental hybrid fusion-fission reactor conceptual design

    International Nuclear Information System (INIS)

    Qiu Lijian; Luan Guishi; Xu Qiang

    1992-03-01

    A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

  9. Transient fission-product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

    1997-12-01

    Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

  10. JENDL-4.0 benchmarking for fission reactor applications

    International Nuclear Information System (INIS)

    Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

    2011-01-01

    Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

  11. The electronuclear cycle: from fission to new reactor systems

    International Nuclear Information System (INIS)

    Belier, G.; Cugnon, J.; Lapoux, V.; Liatard, E.; Porquet, Marie-Genevieve; Rudolf, G.

    2006-09-01

    The Joliot Curie School trains each year, and since 1981, PhD students, post-Doctorates and researchers on scientific breakthroughs performed in a topic related to nuclear physics, in a broad range. These proceedings brings together the 11 lectures given at the 2006 session of Joliot Curie School on the topic of the electronuclear cycle: - Fission: from phenomenology to theory (Berger, J.F.); - Physics of nuclear reactors (Baeten, P.); - Data modeling and evaluation (Bauge, E.; Hilaire, S.); - Measurement of cross sections of interest for minor actinides incineration (Jurado, B.); - Spallation data and modelling for hybrid reactors (Boudard, A.); - Nuclear wastes: overview (Billard, I.); - Long living nuclear wastes transmutation processes and feasibility (Varaine, F.); - Hybrid reactors: recent advances for a demonstrator (Billebaud, A.); - Systems of the future and strategy (David, S.); - Non-nuclear energies (Nifenecker, H.); - Fundamental physics with ultracold neutrons (Protasov, K). The last section is a compilation of abstracts of presentations given by Young searchers' (Young searchers' seminars)

  12. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  13. Photofission observations in reactor environments using selected fission-product yields

    International Nuclear Information System (INIS)

    Gold, R.; Ruddy, F.H.; Roberts, J.H.

    1982-01-01

    A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

  14. Material challenges for the next generation of fission reactor systems

    International Nuclear Information System (INIS)

    Buckthorpe, Derek

    2010-01-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO 2 emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  15. Material challenges for the next generation of fission reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)

    2010-07-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  16. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Science.gov (United States)

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

    2018-01-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  17. Thermoradiation treatment of sewage sludge using reactor waste fission products

    International Nuclear Information System (INIS)

    Reynolds, M.C.; Hagengruber, R.L.; Zuppero, A.C.

    1974-06-01

    The hazards to public health associated with the application of municipal sewage sludge to land usage are reviewed to establish the need for disinfection of sludge prior to its distribution as a fertilizer, especially in the production of food and fodder. The use of ionizing radiation in conjunction with mild heating is shown to be an effective disinfection treatment and an economical one when reactor waste fission products are utilized. A program for researching and experimental demonstration of the process on sludges is also outlined

  18. Gas dynamics models for an oscillating gaseous core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Dam, H. van; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

    1991-01-01

    Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case where a direct energy extraction mechanism (such as magneto-hydrodynamics (MHD)) is not present, increasing density oscillations occur in the gas. Also an estimate is made of the attainable direct energy conversion efficiency, for the case where a direct energy extraction mechanism is present. (author).

  19. Nuclear data requirements for fission reactor neutronics calculations

    International Nuclear Information System (INIS)

    Finck, P.

    1998-01-01

    The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data

  20. Review of the neutron capture process in fission reactors

    International Nuclear Information System (INIS)

    Poenitz, W.P.

    1981-07-01

    The importance of the neutron capture process and the status of the more important cross section data are reviewed. The capture in fertile and fissile nuclei is considered. For thermal reactors the thermal to epithermal capture ratio for 238 U and 232 Th remains a problem though some improvements were made with more recent measurements. The capture cross section of 238 U in the fast energy range remains quite uncertain and a long standing discrepancy for the calculated versus experimental central reaction rate ratio C28/F49 persists. Capture in structural materials, fission product nuclei and the higher actinides is also considered

  1. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  2. A proposed standard on medical isotope production in fission reactors

    International Nuclear Information System (INIS)

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-01-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  3. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    International Nuclear Information System (INIS)

    Was, Gary S.

    2007-01-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

  4. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs

  5. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs.

  6. Muon catalyzed fusion - fission reactor driven by a recirculating beam

    International Nuclear Information System (INIS)

    Eliezer, S.; Tajima, T.; Rosenbluth, M.N.

    1986-01-01

    The recent experimentally inferred value of multiplicity of fusion of deuterium and tritium catalyzed by muons has rekindled interest in its application to reactors. Since the main energy expended is in pion (and consequent muon) productions, we try to minimize the pion loss by magnetically confining pions where they are created. Although it appears at this moment not possible to achieve energy gain by pure fusion, it is possible to gain energy by combining catalyzed fusion with fission blankets. We present two new ideas that improve the muon fusion reactor concept. The first idea is to combine the target, the converter of pions into muons, and the synthesizer into one (the synergetic concept). This is accomplished by injecting a tritium or deuterium beam of 1 GeV/nucleon into DT fuel contained in a magnetic mirror. The confined pions slow down and decay into muons, which are confined in the fuel causing little muon loss. The necessary quantity of tritium to keep the reactor viable has been derived. The second idea is that the beam passing through the target is collected for reuse and recirculated, while the strongly interacted portion of the beam is directed to electronuclear blankets. The present concepts are based on known technologies and on known physical processes and data. 29 refs., 6 figs., 4 tabs

  7. Physics and potentials of fissioning plasmas for space power and propulsion

    Science.gov (United States)

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

    1976-01-01

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

  8. Utilization of fast reactor excess neutrons for burning long-lived fission products

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

  9. Materials compatibility considerations for a fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    DeVan, J.H.; Tortorelli, P.F.

    1983-01-01

    The Tandem Mirror Hybrid Reactor is a fusion reactor concept that incorporates a fission-suppressed breeding blanket for the production of 233 U to be used in conventional fission power reactors. The present paper reports on compatibility considerations related to the blanket design. These considerations include solid-solid interactions and liquid metal corrosion. Potential problems are discussed relative to the reference blanket operating temperature (490 0 C) and the recycling time of breeding materials (<1 year)

  10. An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

  11. Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

    Science.gov (United States)

    Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

    2017-12-01

    A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

  12. Phase 1 space fission propulsion system testing and development progress

    International Nuclear Information System (INIS)

    Van Dyke, Melissa; Houts, Mike; Godfroy, Tom; Dickens, Ricky; Poston, David; Kapernick, Rick; Reid, Bob; Salvail, Pat; Ring, Peter

    2002-01-01

    Successful development of space fission systems requires an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system performance and lifetime can be attained through a series of non-nuclear tests. The Safe Affordable Fission Engine (SAFE) test series, whose ultimate goal is the demonstration of a 300 kW flight configuration system, has demonstrated that realistic testing can be performed using non-nuclear methods. This test series, carried out in collaboration with other NASA centers, other government agencies, industry, and universities, successfully completed a testing program with a 30 kWt core. Stirling engine, and ion engine configuration. Additionally, a 100 kWt core is in fabrication and appropriate test facilities are being reconfigured. This paper describes the current SAFE non-nuclear tests, which includes test article descriptions, test results and conclusions, and future test plans

  13. The role of nuclear reactors in space exploration and development

    Energy Technology Data Exchange (ETDEWEB)

    Lipinski, R.J.

    2000-07-01

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

  14. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

  15. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  16. Some safety studies for conceptual fusion--fission hybrid reactors. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Okrent, D.

    1978-07-01

    The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

  17. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  18. Unique features of space reactors

    International Nuclear Information System (INIS)

    Buden, D.

    1990-01-01

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

  19. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  20. Study of advanced fission power reactor development for the United States. Volume I

    International Nuclear Information System (INIS)

    1976-01-01

    This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

  1. Non Nuclear Testing of Reactor Systems In The Early Flight Fission Test Facilities (EFF-TF)

    International Nuclear Information System (INIS)

    Van Dyke, Melissa; Martin, James

    2004-01-01

    The Early Flight Fission-Test Facility (EFF-TF) can assist in the design and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are 'non-nuclear' in nature (e.g. system's nuclear operations are understood). For many systems, thermal simulators can be used to closely mimic fission heat deposition. Axial power profile, radial power profile, and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other Nasa centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004. (authors)

  2. Energy distribution of antineutrinos originating from the decay of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Rudstam, G.; Aleklett, K.

    1979-01-01

    The energy spectrum of antineutrinos around a nuclear reactor has been derived by summing contributions from individual fission products. The resulting spectrum is weaker at energies above approx. 8 MeV than earlier published antineutrino spectra. The reason may be connected to the strong feeding of high-lying daughter states in the beta decay of fission products with high disintegration energies

  3. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  4. Comparative energetics of three fusion-fission symbiotic nuclear reactor systems

    International Nuclear Information System (INIS)

    Gordon, C.W.; Harms, A.A.

    1975-01-01

    The energetics of three symbiotic fusion-fission nuclear reactor concepts are investigated. The fuel and power balances are considered for various values of systems parameters. The results from this analysis suggest that symbiotic fusion-fission systems are advantageous from the standpoint of economy and resource utilization. (Auth.)

  5. Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

    International Nuclear Information System (INIS)

    Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

    1994-01-01

    To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

  6. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  7. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  8. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    Energy Technology Data Exchange (ETDEWEB)

    Asner, D.M.; Burns, K.; Campbell, L.W.; Greenfield, B.; Kos, M.S., E-mail: markskos@gmail.com; Orrell, J.L.; Schram, M.; VanDevender, B.; Wood, L.S.; Wootan, D.W.

    2015-03-11

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  9. Transport and release of fission products during nuclear reactor accident

    International Nuclear Information System (INIS)

    Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

    1984-01-01

    This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

  10. Cerenkov Detectors for Fission Product Monitoring in Reactor Coolant Water

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1967-09-15

    The expected properties of Cerenkov detectors when used for fission product monitoring in water cooled reactors and test loops are discussed from the point of view of the knowledge of the sensitivity of these detectors to some beta emitting isotopes. The basic theory for calculation of the detector response is presented, taking the optical transmission in the sample container and the properties of the photomultiplier tube into account. Special attention is paid to the energy resolution of this type of Cerenkov detector. For the design of practical detectors the results from several investigations of various window and reflector materials are given, and the selection of photomultiplier tubes is briefly discussed. In the case of optical reflectors and photomultiplier tubes reference is made to two previous reports by the author. The influence of the size and geometry of the sample container on the energy resolution follows from a separate investigation, as well as the relative merits of sample containers with transparent inner walls. Provided that the energy resolution of the Cerenkov detector is sufficiently high, there are several reasons for using this detector type for failed-fuel-element detection. It seems possible to attain the desired energy resolution by careful detector design.

  11. Advanced nuclear fuel production by using fission-fusion hybrid reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

    1993-01-01

    Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

  12. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  13. Determination of the fission coefficients in thermal nuclear reactors for antineutrino detection

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Lenilson M. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Cabral, Ronaldo G., E-mail: rgcabral@ime.eb.b [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Anjos, Joao C.C. dos, E-mail: janjos@cbpf.b [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil). Dept. GLN - G

    2011-07-01

    The nuclear reactors in operation periodically need to change their fuel. It is during this process that these reactors are more vulnerable to occurring of several situations of fuel diversion, thus the monitoring of the nuclear installations is indispensable to avoid events of this nature. Considering this fact, the most promissory technique to be used for the nuclear safeguard for the nonproliferation of nuclear weapons, it is based on the detection and spectroscopy of antineutrino from fissions that occur in the nuclear reactors. The detection and spectroscopy of antineutrino, they both depend on the single contribution for the total number of fission of each actinide in the core reactor, these contributions receive the name of fission coefficients. The goal of this research is to show the computational and mathematical modeling used to determinate these coefficients for PWR reactors. (author)

  14. Role of organic matter in the Proterozoic Oklo natural fission reactors, Gabon, Africa

    International Nuclear Information System (INIS)

    Nagy, B.; Rigali, M.J.; Gauthier-Lafaye, F.; Holliger, P.; Mossman, D.J.; Leventhal, J.S.

    1993-01-01

    Of the sixteen known Oklo and the Bangombe natural fission reactors (hydrothermally altered elastic sedimentary rocks that contain abundant uraninite and authigenic clay minerals), reactors 1 to 6 at Oklo contain only traces of organic matter, but the others are rich in organic substances. Reactors 7 to 9 are the subjects of this study. These organic-rich reactors may serve as time-tested analogues for anthropogenic nuclear-waste containment strategies. Organic matter helped to concentrate quantities of uranium sufficient to initiate the nuclear chain reactions. Liquid bitumen was generated from organic matter by hydrothermal reactions during nuclear criticality. The bitumen soon became a solid, consisting of polycyclic aromatic hydrocarbons and an intimate mixture of cryptocrystalline graphite, which enclosed and immobilized uraninite and the fission-generated isotopes entrapped in uraninite. This mechanism prevented major loss of uranium and fission products from the natural nuclear reactors for 1.2 b.y. 24 refs., 4 figs

  15. System model for analysis of the mirror fusion-fission reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Carlson, G.A.

    1977-01-01

    This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

  16. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  17. 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)

  18. 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)

  19. Fission Fragment Yield Data in Support of Advanced Reactor Technology

    Energy Technology Data Exchange (ETDEWEB)

    Hecht, Adam [Univ. of New Mexico, Albuquerque, NM (United States)

    2017-11-21

    Within the 3 year POP we propose to continue to test and further develop the fission spectrometers, to do development tests and full data acquisition run at the national laboratory neutron beam facilities, to measure correlated fission fragment yields at low neutron energies with 235 U fission targets, and make these data available to the nuclear community. The spectrometer development will be both on the university based r\\prototype and on the National Laboratory Spectrometer, and measurements will be performed with both. Over the longer time frame of the collaboration, we will take data over a range of low energies, and use other fission targets available to the laboratory. We will gather energy specific fragment distributions and reaction cross sections. We will further develop the data acquisition capabilities to take correlated fission fragment'gamma ray/neurton data, all on an event-by-event basis. This really is an enabling technology.

  20. Workshop summaries for the third US/USSR symposium on fusion-fission reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1979-07-01

    Workshop summaries on topics related to the near-term development requirements for fusion-fission (hybrid) reactors are presented. The summary topics are as follows: (1) external factors, (2) plasma engineering, (3) ICF hybrid reactors, (4) blanket design, (5) materials and tritium, and (6) blanket engineering development requirements

  1. 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

  2. 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

  3. A Review of Previous Research in Direct Energy Conversion Fission Reactors

    International Nuclear Information System (INIS)

    DUONG, HENRY; POLANSKY, GARY F.; SANDERS, THOMAS L.; SIEGEL, MALCOLM D.

    1999-01-01

    From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this

  4. AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

  5. Fusion--fission hybrid reactors: a capsule introduction

    International Nuclear Information System (INIS)

    Holdren, J.P.

    1977-01-01

    A short introduction to fusion-fission hybrid systems is provided touching on (a) basic technological characteristics; (b) potential applications; (c) relevance of environmental considerations in the development rationale for hybrids. References to the more technical literature are supplied

  6. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  7. Role of fission-reactor-testing capabilities in the development of fusion technology

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

    1981-01-01

    Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

  8. Design of a Mechanical NaK Pump for Fission Space Power

    Science.gov (United States)

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

    2011-01-01

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

  9. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    International Nuclear Information System (INIS)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F.

    2009-01-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U 235 (typically Pu 242 , Np 237 , U 238 , Th 232 ). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  10. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F. [CEA, DEN, Dosimetry Command Control and Instrumentation Laboratory, F-13109 Saint-Paul-lez-Durance (France)

    2009-07-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U{sup 235} (typically Pu{sup 242}, Np{sup 237}, U{sup 238}, Th{sup 232}). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  11. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    Klein, W.

    1987-01-01

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m -2 . Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN 2 , SiC 2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  12. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Lewis, B.J.

    1983-01-01

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO 2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO 2 (1.56 x 10 -10 to 7.30 x 10 -9 s -1 ), as well as escape rate constants (7.85 x 10 -7 to 3.44 x 10 -5 s -1 ) and diffusion coefficients (3.39 x 10 -5 to 4.88 x 10 -2 cm 2 /s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  13. 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

  14. 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

  15. Reference and standard benchmark field consensus fission yields for U.S. reactor dosimetry programs

    International Nuclear Information System (INIS)

    Gilliam, D.M.; Helmer, R.G.; Greenwood, R.C.; Rogers, J.W.; Heinrich, R.R.; Popek, R.J.; Kellogg, L.S.; Lippincott, E.P.; Hansen, G.E.; Zimmer, W.H.

    1977-01-01

    Measured fission product yields are reported for three benchmark neutron fields--the BIG-10 fast critical assembly at Los Alamos, the CFRMF fast neutron cavity at INEL, and the thermal column of the NBS Research Reactor. These measurements were carried out by participants in the Interlaboratory LMFBR Reaction Rates (ILRR) program. Fission product generation rates were determined by post-irradiation analysis of gamma-ray emission from fission activation foils. The gamma counting was performed by Ge(Li) spectrometry at INEL, ANL, and HEDL; the sample sent to INEL was also analyzed by NaI(Tl) spectrometry for Ba-140 content. The fission rates were determined by means of the NBS Double Fission Ionization Chamber using thin deposits of each of the fissionable isotopes. Four fissionable isotopes were included in the fast neutron field measurements; these were U-235, U-238, Pu-239, and Np-237. Only U-235 was included in the thermal neutron yield measurements. For the fast neutron fields, consensus yields were determined for three fission product isotopes--Zr-95, Ru-103, and Ba-140. For these fission product isotopes, a separately activated foil was analyzed by each of the three gamma counting laboratories. The experimental standard deviation of the three independent results was typically +- 1.5%. For the thermal neutron field, a consensus value for the Cs-137 yield was also obtained. Subsidiary fission yields are also reported for other isotopes which were studied less intensively (usually by only one of the participating laboratories). Comparisons with EBR-II fast reactor yields from destructive analysis and with ENDF/B recommended values are given

  16. Reaction Rate Benchmark Experiments with Miniature Fission Chambers at the Slovenian TRIGA Mark II Reactor

    Science.gov (United States)

    Štancar, Žiga; Kaiba, Tanja; Snoj, Luka; Barbot, Loïc; Destouches, Christophe; Fourmentel, Damien; Villard, Jean-François AD(; )

    2018-01-01

    A series of fission rate profile measurements with miniature fission chambers, developed by the Commisariat á l'énergie atomique et auxénergies alternatives, were performed at the Jožef Stefan Institute's TRIGA research reactor. Two types of fission chambers with different fissionable coating (235U and 238U) were used to perform axial fission rate profile measurements at various radial positions and several control rod configurations. The experimental campaign was supported by an extensive set of computations, based on a validated Monte Carlo computational model of the TRIGA reactor. The computing effort included neutron transport calculations to support the planning and design of the experiments as well as calculations to aid the evaluation of experimental and computational uncertainties and major biases. The evaluation of uncertainties was performed by employing various types of sensitivity analyses such as experimental parameter perturbation and core reaction rate gradient calculations. It has been found that the experimental uncertainty of the measurements is sufficiently low, i.e. the total relative fission rate uncertainty being approximately 5 %, in order for the experiments to serve as benchmark experiments for validation of fission rate profiles. The effect of the neutron flux redistribution due to the control rod movement was studied by performing measurements and calculations of fission rates and fission chamber responses in different axial and radial positions at different control rod configurations. It was confirmed that the control rod movement affects the position of the maximum in the axial fission rate distribution, as well as the height of the local maxima. The optimal detector position, in which the redistributions would have minimum effect on its signal, was determined.

  17. Solid State Track Recorder fission rate measurements in low power light water reactor pressure vessel mockups

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Kellogg, L.S.

    1985-01-01

    The results of extensive SSTR measurements made at the Pool Critical Assembly (PCA) facility at Oak Ridge National Laboratory have been reported previously. Measurements were made at key locations in PCA which is an idealized mockup of the water gap, thermal shield, pressure vessel geometry of a light water reactor. Recently, additional SSTR fission rate measurements have been carried out for 237-Np, 238-U, and 235-U in key locations in the NESTOR Shielding and Dosimetry Improvement Program (NESDIP) mockup facility located at Winfrith, England. NESDIP is a replica of the PCA facility, and comparisons will be made between PCA and NESDIP measurements. The results of measurements made at the engineering mockup at the VENUS critical assembly at CEN/SCK, Mol, Belgium will also be reported. Measurements were made at selected radial and azimuthal locations in VENUS, which models the in-core and near-core regions of a pressurized water reactor. Comparisons of absolute SSTR fission rates with absolute fission rates made with the Mol miniature fission chamber will be reported. Absolute fission rate comparisons have also been made between the NBS fission chamber, radiometric fission foils, and SSTRs, and these results will be summarized

  18. Preparation of a primary target for the production of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

    1976-01-01

    A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

  19. LOFC fission product release and circulating activity calculations for gas-cooled reactors

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

    1977-01-01

    The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

  20. Preconceptual design and analysis of a solid-breeder blanket test in an existing fission reactor

    International Nuclear Information System (INIS)

    Deis, G.A.; Hsu, P.Y.; Watts, K.D.

    1983-01-01

    Preconceptual design and analysis have been performed to examine the capabilities of a proposed fission-based test of a water-cooled Li 2 O blanket concept. The mechanical configuration of the test piece is designed to simulate a unit cell of a breeder-outside-tube concept. This test piece will be placed in a fission test reactor, which provides an environment similar to that in a fusion reactor. The neutron/gamma flux from the reactor produces prototypical power density, tritium production rates, and operating temperatures and stresses. Steady-state tritium recovery from the test piece can be attained in short-duration (5-to-6-day) tests. The capabilities of this test indicate that fission-based testing can provide important near-term engineering information to support the development of fusion technology

  1. New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ma, X.B., E-mail: maxb@ncepu.edu.cn; Qiu, R.M.; Chen, Y.X.

    2017-02-15

    Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between {sup 235}U and {sup 239}Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller. - Highlights: • The covariance coefficients between isotopes vs reactor burnup may change its sign because of two opposite effects. • The relation between fission fraction uncertainty and atomic density are first studied. • A new MC-based method of evaluating the covariance coefficients between isotopes was proposed.

  2. Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

    International Nuclear Information System (INIS)

    Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

    1980-01-01

    It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited

  3. Progress on the conceptual design of a mirror hybrid fusion--fission reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1975-01-01

    A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

  4. Present status of space nuclear reactor

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

  5. Radiation effects in fuel materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.

    1983-01-01

    Physical and chemical changes that occur in fuel materials during fission are described. Emphasis is placed on the fuels used today, or those foreseen for the future, hence oxides and carbides of uranium and plutonium. Examples are given to illustrate the most interesting neutron effects. (author)

  6. Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

    Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

  7. Use of dwell time concept in fission product inventory assessment for CANDU reactors

    International Nuclear Information System (INIS)

    Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

    2003-01-01

    A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

  8. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

  9. Synergies in the design and development of fusion and generation IV fission reactors

    International Nuclear Information System (INIS)

    Bogusch, E.; Carre, F.; Knebel, J.; Aoto, K.

    2007-01-01

    Future fusion reactors or systems and Generation IV fission reactors are designed and developed in worldwide programmes mostly involving the same partners to investigate and assess their potential for realisation and contribution to meet the future energy needs beyond 2030. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core face similar design issues and development needs. Therefore the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactors or systems will be designed for helium and liquid metal cooling and higher temperatures similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches might create synergistic design and development programmes. Therefore an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in support of common technologies. (orig.)

  10. Synergies in the design and development of fusion and generation IV fission reactors

    International Nuclear Information System (INIS)

    Bogusch, E.; Carre, F.; Knebel, J.U.; Aoto, K.

    2008-01-01

    Future fusion reactor and Generation IV fission reactor systems are designed and developed in worldwide programmes to investigate and assess their potential for realisation and contribution to the future energy needs beyond 2030 mostly involving the same partners. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except for the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core, face similar design issues and development needs. Therefore, the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactor systems will be designed for high-temperature helium and liquid metal cooling but also water including supercritical water and molten salt similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches can create synergistic design and development programmes. Therefore, an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in

  11. DIRECT ENERGY CONVERSION (DEC) FISSION REACTORS - A U.S. NERI PROJECT

    International Nuclear Information System (INIS)

    Beller, D.; Polansky, G.

    2000-01-01

    The direct conversion of the electrical energy of charged fission fragments was examined early in the nuclear reactor era, and the first theoretical treatment appeared in the literature in 1957. Most of the experiments conducted during the next ten years to investigate fission fragment direct energy conversion (DEC) were for understanding the nature and control of the charged particles. These experiments verified fundamental physics and identified a number of specific problem areas, but also demonstrated a number of technical challenges that limited DEC performance. Because DEC was insufficient for practical applications, by the late 1960s most R and D ceased in the US. Sporadic interest in the concept appears in the literature until this day, but there have been no recent programs to develop the technology. This has changed with the Nuclear Energy Research Initiative that was funded by the U.S. Congress in 1999. Most of the previous concepts were based on a fission electric cell known as a triode, where a central cathode is coated with a thin layer of nuclear fuel. A fission fragment that leaves the cathode with high kinetic energy and a large positive charge is decelerated as it approaches the anode by a charge differential of several million volts, it then deposits its charge in the anode after its kinetic energy is exhausted. Large numbers of low energy electrons leave the cathode with each fission fragment; they are suppressed by negatively biased on grid wires or by magnetic fields. Other concepts include magnetic collimators and quasi-direct magnetohydrodynamic generation (steady flow or pulsed). We present the basic principles of DEC fission reactors, review the previous research, discuss problem areas in detail and identify technological developments of the last 30 years relevant to overcoming these obstacles. A prognosis for future development of direct energy conversion fission reactors will be presented

  12. Release of radioactive fission products from BN-600 reactor untight fuel elements

    International Nuclear Information System (INIS)

    Osipov, S.L.; Tsikunov, A.G.; Lisitsin, E.C.

    1996-01-01

    The experimental data on the release of radioactive fission products from BN-600 reactor untight fuel elements are given in the report. Various groups of radionuclides: inert gases Xe, Kr, volatile Cs, J, non-volatile Nb, and La are considered. The results of calculation-experimental study of transfer and distribution of radionuclides in the reactor primary circuit, gas system and sodium coolant are considered. It is shown that some complex radioactivity transfer processes can be described by simple mathematical models. (author)

  13. Markets for reactor-produced non-fission radioisotopes

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1995-01-01

    Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included

  14. Investigations of the natural fission reactor program. Progress report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Cowan, G.A.; Norris, A.E.

    1978-10-01

    The U.S. study of the Oklo natural reactor began in 1973 with the principal objectives of understanding the processes that produced the reactor and that led to the retention of many of its products. Major facets of the program have been the chemical separation and mass spectrometric analysis of the reactor components and products, the petrological and mineralogical examination of samples taken from the reactor zones, and an interdisciplinary modeling of possible processes consistent with reactor physics, geophysics, and geochemistry. Most of the past work has been on samples taken within the reactor zones. Presently, these studies give greater emphasis to the measurement of mobile products in additional suites of samples collected peripherally and ''downstream'' from the reactor zones. This report summarizes the current status of research and the views of U.S. investigators, with particular reference to the extensive work of the French scientists, concerning the main features of the Oklo natural fission reactor. Also mentioned briefly is the U.S. search for natural fission reactors at other locations

  15. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    NARCIS (Netherlands)

    Capelli, E.; Beneš, O.; Konings, R.J.M.

    2018-01-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all

  16. A method for measuring power signal background and source strength in a fission reactor

    International Nuclear Information System (INIS)

    Baers, B.; Kall, L.; Visuri, P.

    1977-01-01

    Theory and experimental verification of a novel method for measuring power signal bias and source strength in a fission reactor are reported. A minicomputer was applied in the measurements. The method is an extension of the inverse kinetics method presented by Mogilner et al. (Auth.)

  17. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  18. Thermochemical data for reactor materials and fission products: The ECN database

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1993-02-01

    The activities of the authors regarding the compilation of a database of thermochemical properties for reactor materials and fission products is reviewed. The evaluation procedures and techniques are outlined and examples are given. In addition, examples of the use of thermochemical data for the application in the field of Nuclear Technology are given. (orig.)

  19. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  20. Assessment of the high temperature fission chamber technology for the French fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Geslot, B.; Domenech, T.; Normand, S. [Commissariat a l' Energie Atomique, CEA (France)

    2011-07-01

    High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

  1. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  2. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    International Nuclear Information System (INIS)

    Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-01-01

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  3. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  4. Ternary fission

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Since its discovery in 1946, light (charged) particle accompanied fission (ternary fission) has been extensively studied, for spontaneous as well as for induced fission reactions. The reason for this interest was twofold: the ternary particles being emitted in space and time close to the scission point were expected to supply information on the scission point configuration and the ternary fission process was an important source of helium, tritium, and hydrogen production in nuclear reactors, for which data were requested by the nuclear industry. Significant experimental progress has been realized with the advent of high-resolution detectors, powerful multiparameter data acquisition systems, and intense neutron and photon beams. As far as theory is concerned, the trajectory calculations (in which scission point parameters are deduced from the experimental observations) have been very much improved. An attempt was made to explain ternary particle emission in terms of a Plateau-Rayleigh hydrodynamical instability of a relatively long cylindrical neck or cylindrical nucleus. New results have also been obtained on the so-called open-quotes trueclose quotes ternary fission (fission in three about-equal fragments). The spontaneous emission of charged particles has also clearly been demonstrated in recent years. This chapter discusses the main characteristics of ternary fission, theoretical models, light particle emission probabilities, the dependence of the emission probabilities on experimental variables, light particle energy distributions, light particle angular distributions, correlations between light particle accompanied fission observables, open-quotes trueclose quotes ternary fission, and spontaneous emission of heavy ions. 143 refs., 18 figs., 8 tabs

  5. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  6. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

  7. The temperature distribution in a gas core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).

  8. The temperature distribution in a gas core fission reactor

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C.; Kistemaker, J.; Boersma-Klein, W.; Vitalis, F.

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)

  9. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    International Nuclear Information System (INIS)

    Wang Xinhua; Guo Haiping; Mou Yunfeng; Zheng Pu; Liu Rong; Yang Xiaofei; Yang Jian

    2013-01-01

    A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D + beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three-dimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α, β) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors. (authors)

  10. Fusion-fission hybrid as an alternative to the fast breeder reactor

    International Nuclear Information System (INIS)

    Barrett, R.J.; Hardie, R.W.

    1980-09-01

    This report compares the fusion-fission hybrid on the plutonium cycle with the classical fast breeder reactor (FBR) cycle as a long-term nuclear energy source. For the purpose of comparison, the current light-water reactor once-through (LWR-OT) cycle was also analyzed. The methods and models used in this study were developed for use in a comparative analysis of conventional nuclear fuel cycles. Assessment areas considered in this study include economics, energy balance, proliferation resistance, technological status, public safety, and commercial viability. In every case the characteristics of all fuel cycle facilities were accounted for, rather than just those of the reactor

  11. High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

    1993-01-01

    An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ''engineered safety features,'' which, along with the use of high temperature capable materials further enhance its safety characteristics

  12. In-reactor testing of self-powered neutron detectors and miniature fission chambers

    International Nuclear Information System (INIS)

    Duchene, J.; LeMeur, R.; Verdant, R.

    1975-01-01

    The CEA has tested a variety of ''slow'' self-powered neutron detectors with rhodium, silver and vanadium emitters. Currently there are 120 vanadium detectors in the EL4 heavy water reactor. In addition, ''fast'' detectors with cobalt emitters have been tested at Saclay and 50 of these are in reactor. Other studies are concerned with 6 mm diameter miniature fission chambers. Two fast response chambers with argon-nitrogen filling gas became slow during irradiation, but operated to 600 deg C. An argon filled chamber of 4.7 mm diameter, for traversing in core system in pressurized water reactor, has shown satisfactory test results. (author)

  13. 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

  14. Transfer parameters of fission and activation products present in effluents of nuclear power reactors

    International Nuclear Information System (INIS)

    Cancio, D.; Menossi, C.A.; Ciallella, N.R.

    1978-01-01

    The paper presents results of research carried out in Argentina on transfer parameters of fission and activation products which may be present in the effluents of nuclear power reactors. For some nuclides, as Sr-90, Co-137 and I-131, the parameters were obtained by studies of the fallout, from measurements of integrated levels in the environment and in the food chains. Other values are concentration factors derived from laboratory and field experiments. They refer to fish, molluscs, crustaces and fresh water plants, for several fission and activation nuclides. Transfer parameters obtained have been of significant importance for environmental assessments, relating to nuclear installations in Argentina. (author)

  15. Transmutation of fission products in reactors and accelerator-driven systems

    International Nuclear Information System (INIS)

    Janssen, A.J.

    1994-01-01

    Energy flows and mass flows in several scenarios are considered. Economical and safety aspects of the transmutation scenarios are compared. It is difficult to find a sound motivation for the transmutation of fission products with accelerator-driven systems. If there would be any hesitation in transmuting fission products in nuclear reactors, there would be an even stronger hesitation to use accelerator-driven systems, mainly because of their lower energy efficiency and their poor cost effectiveness. The use of accelerator-driven systems could become a 'meaningful' option only if nuclear energy would be banished completely. (orig./HP)

  16. Inventories of radioactive fission products in the core of thermal nuclear reactor

    International Nuclear Information System (INIS)

    Marinkovic, N.

    1977-01-01

    As a part of the analysis concerning radiological consequences of a major LWR accident, inventories of the most significant radioactive nuclides and stable fission gases in the core of a PWR type reactor have been calculated. Calculations were performed by the DELFIN code using nuclide data and neutron flux data earlier obtained by the METHUSELAH code. Comparison with simplified calculation method show that it is quite rough for certain nuclides but the accuracy may be sufficient for safety analysis purposes recalling the inaccuracies in the later parts of fission product transport process (author)

  17. An assessment of fission product data for decay power calculation in fast reactors

    International Nuclear Information System (INIS)

    Sridharan, M.S.; Murthy, K.P.N.

    1987-01-01

    A review of our present capability at IGC, Kalpakkam to predict fission product decay power in fast reactors is presented. This is accomplished by comparing our summation calculations with the calculations of others and the reported experimental measurements. Our calculations are based on Chandy code developed at our Centre. The fission product data base of Chandy is essentially drawn from the yield data compiled by Crouch (1977) and the data on halflives etc. compiled by Tobias (1973). In general, we find good agreement amongst the different calculations (within ±5%) and our calculations also compare well with experimental measurements of AKIAMA et al and MURPHY et al

  18. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  19. Total Absorption Spectroscopy of Fission Fragments Relevant for Reactor Antineutrino Spectra and Decay Heat Calculations

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

    Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.

  20. Investigation of applications for high-power, self-critical fissioning uranium plasma reactors. Final technical report

    International Nuclear Information System (INIS)

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

    1976-09-01

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

  1. Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

    International Nuclear Information System (INIS)

    Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

    2006-01-01

    A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

  2. Fission track dating method: I. Study of neutron flux uniformity in some irradiation positions of IEA-R1 reactor

    International Nuclear Information System (INIS)

    Osorio, A.M.; Hadler, J.C.; Iunes, P.J.; Paulo, S.R. de

    1993-06-01

    In order to use the fission track dating method the flux gradient was verified within the sample holder, in some irradiation positions of the IEA-R1 reactor at IPEN/CNEN, Sao Paulo. The fission track dating method considers only the thermal neutron fission tracks, to subtract the other contributions sample irradiations with a cadmium cover was performed. The neutron flux cadmium influence was studied. (author)

  3. Journey from discovery of nuclear fission to accelerator-driven sub-critical reactor systems (ADS)

    International Nuclear Information System (INIS)

    Kapoor, S.S.

    2005-01-01

    The epoch making discovery of nuclear fission in 1939, which resulted purely from the curiosity driven basic research to understand the atomic and nuclear structure has changed the world forever with the onset of a new era in the history of human civilization. The basic nuclear physics research pursued after the discovery of fission has also been of much relevance in the harnessing of nuclear energy. In the recent years, there is considerable interest towards developing accelerator driven sub-critical reactor systems (ADS) for the incineration of the long-lived spent fuel radioactive waste and for the utilization of thorium fuel for nuclear power generation. In this talk, we discuss important milestones in the journey from discovery of nuclear fission to ADS. (author)

  4. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  5. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and deposition of fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behavior of simulant fission product species such as cesium iodide, cesium hydroxide and tellurium, in terms of their vapor deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO 2 clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapor phase, and specific data using this technique are reported

  6. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

    1997-01-01

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  7. Potentials of fissioning plasmas

    International Nuclear Information System (INIS)

    Karlheinz, Thom.

    1979-01-01

    Successful experiments with the nuclear pumping of lasers have demonstrated that in gaseous medium the kinetic energy of fission fragments can be converted directly into non-equilibrium optical radiation. This confirms the concept that the fissioning medium in a gas-phase nuclear reactor shows an internal structure such as a plasma in nearly thermal equilibrium varying up to a state of extreme-non-equilibrium. The accompanying variations of temperatures, pressure and radiative spectrum suggest wide ranges of applications. For example, in the gas-phase fission reactor concept enriched uranium hexafluoride or an uranium plasma replaces conventional fuel elements and permits operation above the melting point of solid materials. This potential has been motivation for the US National Aeronautics and Space Administration (NASA) to conduct relevant research for high specific impulse propulsion in space. The need to separate the high temperature gaseous fuel from the surfaces of a containing vessel and to protect them against thermal radiation has led to the concept of an externally moderated reactor in which the fissioning gaseous material is suspended by fluid dynamic means and the flow of opaque buffer gas removes the power. The gaseous nuclear fuel can slowly be circulated through the reactor for continuous on-site reprocessing including the annihilation of transuranium actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides can thus be achieved. These characteristics and the unique radiative properties led to the expectation that the gas-phase fission reactor could feature improved safety, safeguarding and economy, in addition to new technologies such as processing, photochemistry and the transmission of power over large distances in space

  8. 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.

  9. Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

    International Nuclear Information System (INIS)

    Adlys, G.; Adliene, D.

    2002-01-01

    Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

  10. Geochemical properties and nuclear chemical characteristics of Oklo natural fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, Hiroshi [Hiroshima Univ., Higashi-Hiroshima (Japan). Faculty of Science

    1997-07-01

    There are six uranium deposits in the Gabonese Republic in the cnetral Africa. `Fission reactor zone`, the fission chain reactions generated about 200 billion years ago, was existed in a part of them. CEA begun geochemical researches of Oklo deposits etc. in 1991. The geochemical and nuclear chemical properties of Oklo were reviewed from the results of researches. Oklo deposits is consisted of main five sedimentary faces such as sandstone (FA), Black Shale formation (FB), mudstone (FC), tuff (FD) and volcaniclastic sandstone (FE) from the bottom on the base rock of granite in the Precambrian era. Uranium is enriched in the upper part of FA layer and the under part of FB layer. {sup 235}U/{sup 238}U, U content, fission proportion, duration time, neutron fluence, temperature, restitution factor of {sup 235}U and epithermal index ({gamma}) were investigated and compared. The geochemical properties of Oklo are as followed: large enrich of uranium, the abundance ratio of {sup 235}U as same as that of enriched uranium, interaction of natural water and small rear earth elements. These factors made casually Oklo fission reactor. (S.Y.)

  11. Space reactors - past, present, and future

    International Nuclear Information System (INIS)

    Buden, D.; Angelo, J.

    1983-01-01

    In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

  12. Molten fluoride mixtures as possible fission reactor fuels

    International Nuclear Information System (INIS)

    Grimes, W.R.

    1978-01-01

    Molten mixtures of fluorides with UF 4 as a component have been used as combined fuel and primary heat transfer agent in experimental high-temperature reactors and have been proposed for use in breeders or converters of 233 U from thorium. Such use places stringent and diverse demands upon the fluid fuel. A brief review of chemical behavior of molten fluorides is given to show some of their strengths and weaknesses for such service

  13. Application of Campbell's MSV method in monitoring of reactor's fission power

    International Nuclear Information System (INIS)

    Stankovic, S.J.; Vukcevic, M.; Loncar, B.; Vasic, A.; Osmokrovic, P.

    2003-01-01

    This paper presents some possibilities of Campbell's MSV (Mean Square Value) method in monitoring the reactor's fission power. Investigation of gamma discrimination compared to neutron component of signal along with change of variance and mean value the detector output signal for a specified range of reactor's fission power (10mW-22W) was carried out. The uncompensated ionization chamber for mixed n- gamma fields was used as detector element. Experimental measurements were performed using digitized MSV method, and obtained results were compared to those obtained by classical measuring chain. The final conclusion is that the order of discrimination in MSV signal processing is about fifty times larger than for classical measuring method (author)

  14. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Science.gov (United States)

    Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

    2016-02-01

    The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  15. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Directory of Open Access Journals (Sweden)

    Wagemans Jan

    2016-01-01

    Full Text Available The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  16. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  17. Major features of a mirror fusion--fast fission hybrid reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1974-01-01

    A conceptual design was made of a fusion-fission reactor. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and sustained by hot neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and is cooled by helium. It was shown how the reactor can be built using essentially present day construction technology and how the uranium bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel of which approximately 1200 kg of plutonium are produced each year along with the approximately 750 MW of electricity. (U.S.)

  18. Feasibility study of a fission supressed blanket for a tandem-mirror hybrid reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Barr, W.L.

    1981-01-01

    A study of fission suppressed blankets for the tandem mirror not only showed such blankets to be feasible but also to be safer than fissioning blankets. Such hybrids could produce enough fissile material to support up to 17 light water reactors of the same nuclear power rating. Beryllium was compared to 7 Li for neutron multiplication; both were considered feasible but the blanket with Li produced 20% less fissile fuel per unit of nuclear power in the reactor. The beryllium resource, while possibly being too small for extensive pure fusion application, would be adequate (with carefully planned industrial expansion) for the hybrid because of the large support ratio, and hence few hybrids required. Radiation damage and coatings for beryllium remain issues to be resolved by further study and experimentation. Molten salt reprocessing was compared to aqueous solution reprocessing

  19. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    International Nuclear Information System (INIS)

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  20. Japanese list of requests for neutron nuclear data for fission reactors

    International Nuclear Information System (INIS)

    Igarasi, Sin-iti; Asami, Tetsuo

    1977-05-01

    Requests for neutron nuclear data for fission reactors are presented. These are screened by a WRENDA Working Group of Japanese Nuclear Data Committee and submitted to WRENDA 76/77. This report includes 163 requests of which 55 requests are newly registered in the WRENDA. Three requests of the previous list are withdrawn. This activity is a part of the international cooperation with CCDN, NEANDC and INDC. (auth.)

  1. Pressure due to fission gases in a fuel element circulating in a reactor

    International Nuclear Information System (INIS)

    Fonteray, Jean

    1965-01-01

    This document states calculation hypotheses and methods used to assess pressures due to fission gases in a fuel element moving in a reactor channel in the reverse direction with respect to the cooling fluid. The calculation comprises the calculation of the temperature in the fuel rod, of the reduced diffusion coefficient, of the diffused gas fraction, of the pressure. The appendix describes the use of the SPM 076 software: input data, output results, computing time [fr

  2. Feasibility study of a fission-suppressed tandem-mirror hybrid reactor

    International Nuclear Information System (INIS)

    Lee, J.D.; Moir, R.W.; Barr, W.L.

    1982-04-01

    Results of a conceptual design study of a U-233 producing fusion breeder consisting of a tandem mirror fusion device and two types of fission-suppressed blankets are presented. The majority of the study was devoted to the conceptual design and evaluation of the two blankets. However, studies in the areas of fusion engineering, reactor safety, fuel reprocessing, other fuel cycle issues, economics, and deployment were also performed

  3. Resuspension of fission products during severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    Borkowski, R.; Bunz, H.; Schoeck, W.

    1986-05-01

    This report investigates the influence of resuspension phenomena on the overall radiological source term of core melt accidents in a pressurized water reactor. A review of the existing literature is given and the literature data are applied to calculations of the source term. A large scatter in the existing data was found. Depending on the scenario and on the data set chosen for the calculations the relative influence of resuspended fission products on the source term ranges from dominant to negligible. (orig.) [de

  4. Space reactors, a prospective for the future

    International Nuclear Information System (INIS)

    Wahlquist, E.; Voss, S.S.

    1989-01-01

    The power requirements for future space missions are increasing and alternate power systems will be required to meet these needs. Therefore, in the early 1980's a tri-agency space reactor program, the SP-100, was initiated that is capable of meeting the higher power requirements. To understand the current space reactor program, it is important to review it in the context of past space nuclear programs - including radioisotopes, nuclear rockets and reactors. Initial effort on these programs began in the mid-1950's. Radioisotope generators have been flown on a variety of missions and are continuing to be used. The space reactor and nuclear rocket programs were technically successful but were both terminated in 1973. The current SP-100 program builds on those earlier programs

  5. Simulation of fusion first-wall environment in a fission reactor

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

    1982-01-01

    A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

  6. Conceptual design study of Hyb-WT as fusion–fission hybrid reactor for waste transmutation

    International Nuclear Information System (INIS)

    Siddique, Muhammad Tariq; Kim, Myung Hyun

    2014-01-01

    Highlights: • Conceptual design study of fusion-fission hybrid reactor for waste transmutation. • MCNPX and MONTEBURNS are compared for transmutation performance of WT-Hyb. • Detailed neutronic performance of final optimized Hyb-WT design is analyzed. • A new tube-in-duct core design is implemented and compared with pin type design. • Study shows many aspects of hybrid reactor even though scope was limited to neutronic analysis. - Abstract: This study proposes a conceptual design of a hybrid reactor for waste transmutation (Hyb-WT). The design of Hyb-WT is based on a low-power tokamak (less than 150 MWt) and an annular ring-shaped reactor core with metal fuel (TRU 60 w/o, Zr 40 w/o) and a fission product (FP) zone. The computational code systems MONTEBURNS and MCNPX2.6 are investigated for their suitability in evaluating the performance of Hyb-WT. The overall design performance of the proposed reactor is determined by considering pin-type and tube-in-duct core designs. The objective of such consideration is to explore the possibilities for enhanced transmutation with reduced wall loading from fusion neutrons and reduced transuranic (TRU) inventory. TRU and FP depletion is analyzed by calculating waste transmutation ratio, mass burned per full power year (in units of kg/fpy), and support ratio. The radio toxicity analysis of TRUs and FPs is performed by calculating the percentage of toxicity reduction in TRU and FP over a burn cycle

  7. Source driven breeding fission power reactors and the nuclear energy strategy

    International Nuclear Information System (INIS)

    Greenspan, E.

    The nuclear energy economy is facing severe difficulties associated with low utilization of uranium resources, safety, non-proliferation and environmental issues. Energy policy makers face the dilemma: commercialize LMFBRs immediately with the risk of negative economical, proliferation or other consequences, or continue with R and D programs that will provide the information needed for sounder decisions, but now taking the risk of running out of economically exploitable uranium ore resources. The development of hybrid reactors can provide an assurance against the latter risk and offers many interesting new options for the nuclear energy strategy. Being based on the technology of LWRs and HWRs, Light Water Hybrid Reactors (LWHR) provide a most natural link between the fission reactor technology of the present and the fusion power technology of the future. The investment in their development in excess of that required for the development of fusion power reactors is expected to be relatively small, thus making the development of LWHRs potentially a high benefit-to-cost ratio program. It is recommended that the fission and fusion communities will cooperate in hybrids R and D programs aimed at assessing the technological and economical viability of hybrid reactors as reliably and soon as possible. (author)

  8. Phase 1 space fission propulsion system design considerations

    International Nuclear Information System (INIS)

    Houts, Mike; Van Dyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert

    2002-01-01

    Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a 'Phase 1' fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system

  9. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  10. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  11. Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

    International Nuclear Information System (INIS)

    Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

    1978-01-01

    The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

  12. Conceptual design of a hybrid fusion-fission reactor with intrinsic safety and optimized energy productivity

    International Nuclear Information System (INIS)

    Talebi, Hosein; Sadat Kiai, S.M.

    2017-01-01

    Highlights: • Designing a high yield and feasible Dense Plasma Focus for driving the reactor. • Presenting a structural method to design the dual layer cylindrical blankets. • Finding, the blanket production energy, in terms of its geometrical and material parameters. • Designing a subcritical blanket with optimization of energy amplification in detail. - Abstract: A hybrid fission-fusion reactor with a Dense Plasma Focus (DPF) as a fusion core and the dual layer fissionable blanket as the energy multiplier were conceptually designed. A cylindrical DPF, energized by a 200 kJ bank energy, is considered to produce fusion neutron, and these neutrons drive the subcritical fission in the surrounding blankets. The emphasis has been placed on the safety and energy production with considering technical and economical limitations. Therefore, the k eff-t of the dual cylindrical blanket was defined and mathematically, specified. By applying the safety criterion (k eff-t ≤ 0.95), the geometrical and material parameters of the blanket optimizing the energy amplification were obtained. Finally, MCNPX code has been used to determine the detailed dimensions of the blankets and fuel rods.

  13. Direct energy conversion in fission reactors: A U.S. NERI project

    International Nuclear Information System (INIS)

    Slutz, Stephen A.; Seidel, David B.; Polansky, Gary F.; Rochau, Gary E.; Lipinski, Ronald J.; Besenbruch, G.; Brown, L.C.; Parish, T.A.; Anghaie, S.; Beller, D.E.

    2000-01-01

    In principle, the energy released by a fission can be converted directly into electricity by using the charged fission fragments. The first theoretical treatment of direct energy conversion (DEC) appeared in the literature in 1957. Experiments were conducted over the next ten years, which identified a number of problem areas. Research declined by the late 1960's due to technical challenges that limited performance. Under the Nuclear Energy Research Initiative the authors are determining if these technical challenges can be overcome with todays technology. The authors present the basic principles of DEC reactors, review previous research, discuss problem areas in detail, and identify technological developments of the last 30 years that can overcome these obstacles. As an example, the fission electric cell must be insulated to avoid electrons crossing the cell. This insulation could be provided by a magnetic field as attempted in the early experiments. However, from work on magnetically insulated ion diodes they know how to significantly improve the field geometry. Finally, a prognosis for future development of DEC reactors will be presented

  14. In-reactor measurements of thermo mechanical behaviour and fission gas release of water reactor fuel

    International Nuclear Information System (INIS)

    Kolstad, E.; Vitanza, C.

    1983-01-01

    the fuel performance during and after a power ramp can be investigated by direct in-pile measurements related to the thermal, mechanical and fission gas release behaviour. The thermal response is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by the simultaneous liberation of fission gases. The thermal feedback effect is also being separately studied out-of-pile in a specially designed rod where the fission gas release is simulated by injecting xenon in known quantities at different axial positions within the rod. Investigations on the mechanical behaviour are based on axial and diametral cladding deformation measurements. This enables the determination of the amount of local cladding strain and ridging during ramping, the extent of relaxation during the holding time and the amount of residual (plastic) deformation. Gap width measurements are also performed in operating fuel rods using a cladding deflection technique. Fission gas release data are obtained, besides from post-irradiation puncturing, by continuous measurements of the rod internal pressure. This type of measurement leads to the description of the kinetics of the fission gas release process at different powers. The data tend to indicate that the time-dependent release can be reasonably well described by simple diffusion. The paper describes measuring techniques developed and currently in use in Halden, and presents and discusses selected experimental results obtained during various power ramps and transients. (author)

  15. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  16. 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.

  17. Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

    International Nuclear Information System (INIS)

    Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2014-01-01

    In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  18. Survey on the fusion/fission-hybrid-reactors, a literature review

    International Nuclear Information System (INIS)

    A survey, based on existing literature, of the work being pursued worldwide on fusion - fission (hybrid) reactor systems is presented. Six areas are reviewed: Plasma physics parameters; Blankets concepts; Fuel cycles; Reactor conceptual designs; Safety and environmental problems; System studies and economic perspectives. Attention has been restricted to systems using magnetically confined plasmas, mainly to mirror and Tokamak - type concepts. The aim is to provide sufficient information, even if not exhaustive, on hybrid reactor concepts in order to help understand what may be expected from their possible development and the ways in which hybrids could affect the future energy scenario. Some concluding remarks are made which represent the personal view of the authors only

  19. Energy deposition measurements in fast reactor safety experiments with fission thermocouple detectors

    International Nuclear Information System (INIS)

    Wright, S.A.; Scott, H.L.

    1979-01-01

    The investigation of phenomena occurring in in-pile fast reactor safety experiments requires an accurate measurement of the time dependent energy depositions within the fissile material. At Sandia Laboratories thin-film fission thermocouples are being developed for this purpose. These detectors have high temperature capabilities (400 to 500 0 C), are sodium compatible, and have milli-second time response. A significant advantage of these detectors for use as energy deposition monitors is that they produce an output voltage which is directly dependent on the temperature of a small chip of fissile material within the detectors. However, heat losses within the detector make it necessary to correct the response of the detector to determine the energy deposition. A method of correcting the detector response which uses an inverse convolution procedure has been developed and successfully tested with experimental data obtained in the Sandia Pulse Reactor (SPR-II) and in the Annular Core Research Reactor

  20. RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

    Directory of Open Access Journals (Sweden)

    GO CHIBA

    2014-06-01

    Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  1. Needs and accuracy requirements for fission product nuclear data in the physics design of power reactor cores

    International Nuclear Information System (INIS)

    Rowlands, J.L.

    1978-01-01

    The fission product nuclear data accuracy requirements for fast and thermal reactor core performance predictions were reviewed by Tyror at the Bologna FPND Meeting. The status of the data was assessed at the Meeting and it was concluded that the requirements of thermal reactors were largely met, and the yield data requirements of fast reactors, but not the cross section requirements, were met. However, the World Request List for Nuclear Data (WRENDA) contains a number of requests for fission product capture cross sections in the energy range of interest for thermal reactors. Recent reports indicate that the fast reactor reactivity requirements might have been met by integral measurements made in zero power critical assemblies. However, there are requests for the differential cross sections of the individual isotopes to be determined in addition to the integral data requirements. The fast reactor requirements are reviewed, taking into account some more recent studies of the effects of fission products. The sodium void reactivity effect depends on the fission product cross sections in a different way to the fission product reactivity effect in a normal core. This requirement might call for different types of measurement. There is currently an interest in high burnup fuel cycles and alternative fuel cycles. These might require more accurate fission product data, data for individual isotopes and data for capture products. Recent calculations of the time dependence of fission product reactivity effects show that this is dependent upon the data set used and there are significant uncertainties. Some recent thermal reactor studies on approximations in the treatment of decay chains and the importance of xenon and samarium poisoning are also summarized. (author)

  2. Study on the technical feasibility of Fission-Track dating at two irradiation positions of the RA-6 research reactor

    International Nuclear Information System (INIS)

    Dorval, Eric

    2005-01-01

    The method of Fission-Track dating is based upon the detection of the damage caused by fission fragments from the Uranium contained in geological samples.In order to determine the age of a sample, both the amount of spontaneous fissions occurred and the Uranium concentration must be known.The latter requires the irradiation of the samples inside a reactor with a well-thermalized flux, so that fissions are induced over 235 U targets only. Therefore, the Uranium concentration may be determined.The main inconvenient presented by the irradiation sites at the RA-6 MTR-type reactor is that neutron flux is not completely thermal there, which means that fissions due to epithermal and fast neutrons will not be negligible.In the same way, tracks due to fissions of 238 U and 232 Th will be detected. In order to know the corrections that must be applied to those measurements performed in this reactor, it is necessary to characterize fast flux.Because of it, this laboratory's gamma spectrometry equipment had to be calibrated. After that, several activation detectors were irradiated and results were analyzed. Finally, it was determined that it is feasible to Fission-Track date at the I6 position. However, limitations associated to this method were analyzed for the values of flux measured in the different sites

  3. Migration of U-series radionuclides around the Bangombe natural fission reactor (Gabon)

    International Nuclear Information System (INIS)

    Bros, R.; Yanase, N.; Isobe, H.; Sato, T.; Iida, Y.; Ohnuki, T.; Roos, P.; Holm, E.

    1999-01-01

    The Bangombe natural fission reactors has undergone extensive weathering phenomena and continues to be affected by the penetration of meteoric waters. Hence this system provides a model for studying the stability of spent fuel uraninite and the influence of various rock matrices on the mobilization/retardation of various actinides and fission products. The Bangombe uranium deposit has been investigated by drilling on a grid. Radiochemical analysis by alpha- and gamma-spectroscopy of the obtained rocks show significant disequilibria of the 234 U/ 238 U, 230 Th/ 234 U, and 226 Ra/ 230 Th parent-daughter pairs. In this paper, a conceptual model for spatio/temporal evolution of the Bangombe system is proposed. (J.P.N.)

  4. The study of two, three and four dimensional nonlinear dynamics of nuclear fission reactors and effective parameters on its behaviour

    International Nuclear Information System (INIS)

    Tajik, M.; Ghasemizad, A.

    2008-01-01

    In this research, new physical fission reactor parameters which have very sensitive effects on the qualitative behavior of a reactor, are introduced. Therefore, the two, the nonlinear dynamics of two, three and four dimensional, considering almost the effective parameters are formulated for describing nuclear fission reactor systems. Using both analytical and numerical methods, the stability and instability of the given dynamical equations and the conditions of stability are studied in these systems. We have shown that the two parameters of the mean energy residence time in fuel and coolant and also their ratios have the most qualitative effects on the dynamical behaviour of a typical nuclear fission reactor. Increasing or decreasing of these parameters from a captain limit can lead to stability or un stability in a given system

  5. Planetary Surface Power and Interstellar Propulsion Using Fission Fragment Magnetic Collimator Reactor

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.; Hart, Ron R.; King, Don B.; Rochau, Gary E.

    2006-01-01

    Fission energy can be used directly if the kinetic energy of fission fragments is converted to electricity and/or thrust before turning into heat. The completed US DOE NERI Direct Energy Conversion (DEC) Power Production project indicates that viable DEC systems are possible. The US DOE NERI DEC Proof of Principle project began in October of 2002 with the goal to demonstrate performance principles of DEC systems. One of the emerging DEC concepts is represented by fission fragment magnetic collimator reactors (FFMCR). Safety, simplicity, and high conversion efficiency are the unique advantages offered by these systems. In the FFMCR, the basic energy source is the kinetic energy of fission fragments. Following escape from thin fuel layers, they are captured on magnetic field lines and are directed out of the core and through magnetic collimators to produce electricity and thrust. The exiting flow of energetic fission fragments has a very high specific impulse that allows efficient planetary surface power and interstellar propulsion without carrying any conventional propellant onboard. The objective of this work was to determine technological feasibility of the concept. This objective was accomplished by producing the FFMCR design and by analysis of its performance characteristics. The paper presents the FFMCR concept, describes its development to a technologically feasible level and discusses obtained results. Performed studies offer efficiencies up to 90% and velocities approaching speed of light as potentially achievable. The unmanned 10-tons probe with 1000 MW FFMCR propulsion unit would attain mission velocity of about 2% of the speed of light. If the unit is designed for 4000 MW, then in 10 years the unmanned 10-tons probe would attain mission velocity of about 10% of the speed of light

  6. 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

  7. New generation of reactors for space power

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Buden, D.

    1982-01-01

    Space nuclear reactor power is expected to enable many new space missions that will require several times to several orders of magnitude anything flown in space to date. Power in the 100-kW range may be required in high earth orbit spacecraft and planetary exploration. The technology for this power system range is under development for the Department of Energy with the Los Alamos National Laboratory responsible for the critical components in the nuclear subsystem. The baseline design for this particular nuclear sybsystem technology is described in this paper; additionally, reactor technology is reviewed from previous space power programs, a preliminary assessment is made of technology candidates covering an extended power spectrum, and the status is given of other reactor technologies

  8. Innovative High Temperature Heat Pipes for Spacecraft Nuclear Fission Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA Glenn is examining small fission reactors for future space transportation and surface power applications. The reactors would have an 8 to 15 year design life...

  9. Data sheets of fission product release experiments for light water reactor fuel, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo; Yamamoto, Katsumune; Nakazaki, Chozaburo.

    1979-07-01

    This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO 2 pellets, photographs of parts of fuel rod assemblies for irradiation and the assemblies, operational conditions of JMTR and OWL-1, variations of radioiodine-131 level in the main loop coolant during experimental periods, and representative results of post-irradiation examinations of respective fuel rods. (author)

  10. Process for separately recovering uranium, transuranium elements, and fission products of uranium from atomic reactor fuel

    International Nuclear Information System (INIS)

    Balal, A.L.; Metscher, K.; Muehlig, B.; Reichmuth, C.; Schwarz, B.; Zimen, K.E.

    1976-01-01

    Spent reactor fuel elements are dissolved in dilute nitric acid. After addition of acetic acid as a complexing agent, the nitric acid is partly decomposed and the mixture subjected to electrolysis while a carrier liquid, which may be dilute acetic acid or a dilute mixture of acetic acid and nitric acid is caused to flow in the electric field between the electrodes either against the direction of ion migration or transversely thereto. The ions of uranium, plutonium, and other transuranium elements, and of fission products accumulate in discrete portions of the electrolyte and are separately withdrawn as at least three fractions after one or more stages of electrolysis

  11. The effect of the greek research reactor operating schedule on its fission product inventory

    International Nuclear Information System (INIS)

    Annousis, J.N., Armyriotis, J.S.

    1987-12-01

    A simple method to convert the fission product inventory of the 'Democritos' uous Greek Research Reactor (GRR) corresponding to its continuous operation over a given time interval, into the inventory corresponting to GRR discontinuous but periodic operation of the same total duration, is presented in this paper. Relevant correction factors for 31 radioecologically significant radionuclides of the inventory are given as a function of the number of hours or operation per day, 5 days per week of the GRR, according to its present of possible future operating schedule

  12. The effect of the Greek Research Reactor operating schedule on its fission product inventory

    International Nuclear Information System (INIS)

    ANOUSSIS, J.N.; ARMYRIOTIS, J.S.

    1987-12-01

    Full text:A simple method to convert the fission product inventory of ''Demokritos'' Greek Research Reactor(GRR) corresponding to its continuous operation over a given time interval, into the inventory corresponding to GRR discontinuous but periodic operation of the same total duration, is presented in this paper. Relevant correction factors for 31 radioecologically significant radionuclides of the inventory are given as a function of the number of hours of operation per day, 5 days per week of the GRR, according to its present or possible future operating schedule. (author)

  13. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

    1997-05-01

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  14. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  15. 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

  16. The geo-reactor. A link between nuclear fission and geothermal energy?

    International Nuclear Information System (INIS)

    Degueldre, Claude; Fiorina, Carlo

    2013-01-01

    Recent high-precision isotope analysis data suggests the potential occurrence of a geo-reactor. Specific gas isotopes that could have been generated by binary and ternary fissions were identified in volcano emanations or as soluble/associated species in crystalline rocks and semi-quantitatively evaluated as isotopic ratio or estimated amounts. Presently if it is evident that according to the actinide inventory on the Earth, local potential criticality of the geo-system may have been reached, several questions remain such as why, where and when did a geo-reactor be operational? Even if the hypothesis of a geo-reactor operation in the proto-Earth period should be acceptable, it could be difficult to anticipate that a geo-reactor is still operating today. This could be tested in the future by assessing and reconstructing the system by antineutrino detection and tomography through the Earth. The present paper focuses on the geo-reactor conditions including history, spatial extension and regimes. The discussion based on recent calculations involves investigations on the limits in term of fissile inventory, size and power, based on stratification through the gravitational field and the various features through the inner mantel, the boundary with the core, the external part and the inner-core. the reconstruction allows to formulating that from the history point of view there are possibilities that the geo-reactor reached criticality in a proto-Earth period as a thorium/uranium reactor triggered by an under-layer with heavier actinides. The geo-reactor should be a key component of geothermal energy sources. (author)

  17. Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

    International Nuclear Information System (INIS)

    Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

    2002-01-01

    An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

  18. Space reactors - past, present, and future

    International Nuclear Information System (INIS)

    Buden, D.; Angelo, J.A.

    1983-01-01

    The successful test flights of the Space Shuttle mark the start of a new era--an era of routine manned access into cislunar space. Human technical development at the start of the next Millenium will be highlighted by the creation of Man's extraterrestrial civilization with off-planet expansion of the human resource base. In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

  19. Measurement of fission cross-section of actinides at n_TOF for advanced nuclear reactors

    CERN Document Server

    Calviani, Marco; Montagnoli, G; Mastinu, P

    2009-01-01

    The subject of this thesis is the determination of high accuracy neutron-induced fission cross-sections of various isotopes - all of which radioactive - of interest for emerging nuclear technologies. The measurements had been performed at the CERN neutron time-of-flight facility n TOF. In particular, in this work, fission cross-sections on 233U, the main fissile isotope of the Th/U fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on these isotopes are requested for the feasibility study of innovative nuclear systems (ADS and Generation IV reactors) currently being considered for energy production and radioactive waste transmutation. The measurements have been performed with a high performance Fast Ionization Chamber (FIC), in conjunction with an innovative data acquisition system based on Flash-ADCs. The first step in the analysis has been the reconstruction of the digitized signals, in order to extract the information required for the discrimination between fission fragm...

  20. A study of fission product transport from failed fuel during N reactor postulated accidents

    International Nuclear Information System (INIS)

    Hagrman, D.L.

    1989-09-01

    This report presents a study of fission product transport behavior in N Reactor during a severe accident. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The current report is an extension to a previous interum study that has added an aerosol formation model, replaced an older aerosol deposition model with an improved correlation, and incorporated results of a revised analysis of the process tubes. The LACE LA1 and LA3 tests are used to assess the revised model applied to determine aerosol deposition. The study concludes that a cesium iodide aerosol is likely to form near the downstream end of the process tubes. Transport of most of the released cesium and iodine as well as less volatile material depends on the behavior of this aerosol and the behavior is sensitive to several parameters that are not well known. If the environment is very clean and effluent flow is sufficient to support oxidation of the zircaloy and uranium of the process tubes, almost none of the aerosol deposits in the riser. Reduction of the effluent flow or the presence of high concentrations of aerosols of very low volatile material like zirconium, uranium, or their oxides causes deposition of the fission products in the riser piping. 24 refs., 18 figs., 11 tabs

  1. New ceramics for nuclear industry. Case of fission and fusion reactors

    International Nuclear Information System (INIS)

    Yvars, M.

    1979-10-01

    The ceramics used in the nuclear field are described as is their behaviour under radiation. 1) Power reactors - nuclear fission. Ceramics enter into the fabrication of nuclear fuels: oxides, carbides, uranium or plutonium nitrides or oxy-nitrides. Silicon carbide SiC is used for preparing the fuels of helium cooled high temperature reactors. Its use is foreseen in the design of gas high temperature gas thermal exchangers, as is silicon nitride (Si 3 N 4 ). In the materials for safety or control rods, the intense neutron flows induce nuclear reactions which increase the temperature of the neutron absorbing material. Boron carbide B 4 C, rare earth oxides Ln 2 O 3 , or B 4 C-Cu or B 4 C-Al cermets are employed. Burnable poison materials are formed of Al 2 O 3 -B 4 C or Al 2 O 3 -Ln 2 O 3 cermets. The moderators of thermal neutron reactors are in high purety polycrystalline graphite. For the thermal insulation of reactor vessels and jackets, honeycomb ceramics are used as well as ceramic fibres on an increasing scale (kaolin, alumina and other fibres). 2) fusion reactors (Tokomak). These require refractory materials with a low atomic number. Carbon fibres, boron carbide, some borons (Al B 12 ), silicon nitrides and oxy-nitrides and high density alumina are the substances considered [fr

  2. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  3. Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, RE

    2004-07-15

    The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

  4. The different facilities of the reactor Phenix for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    Coulon, P.; Clerc, R.; Tommasi, J.

    1993-01-01

    During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (authors). 9 figs

  5. 99Tc, Pb and Ru migration around the Oklo natural fission reactors

    International Nuclear Information System (INIS)

    Gancarz, A.; Cowan, G.; Curtis, D.; Maeck, W.

    1980-01-01

    This work demonstrates the utility of the Oklo uranium ore deposit and natural fission reactors as a long time scale analogue for man-made radioactive waste repositories. It has been shown that the ores and nearby rocks were open to the loss and gain of 99 Tc, ruthenium, and lead relative to uranium. Identified regions of element deficiencies and those which are correspondingly enriched are separated by less than 10 meters. However, more extensive sampling is required to define the overall extent of the element migration. Element fractionation took place on at least two vastly different time scales; 99 Tc was fractionated from ruthenium within one million years of the end of reactor criticality. Lead-uranium fractionation has been ongoing for most of the two billion years since the ores were formed. Diffusion loss of lead from host uraninite appears to be an important process in the fractionation of lead from uranium

  6. Transmutation of long-lived fission product (137Cs, 90Sr) by a reactor-accelerator system

    International Nuclear Information System (INIS)

    Toyama, Shin-ichi; Takashita, Hirofumi; Konashi, Kenji; Sasao, Nobuyuki; Sato, Isamu.

    1990-01-01

    The report discusses the transmutation of long-lived fission products by a reactor and accelerator. It is important to take some criteria into consideration in transmutation disposal. To satisfy the criteria, a combined system of a reactor and an accelerator is proposed for the transmutation. An outline of the transmutation reactor and the accelerator is presented. The transmutation reactor has the ability to transmute a large quantity of fission products. However, it is desirable to have a high transmutation rate as well as a large disposal ability. Besides the transmutation property, it is necessary to investigate the physics of the transmutation reactor such as nuclear characteristics and burnup properties in order to obtain the most suitable, high performance core concept. A study on those properties is also presented. A high power accelerator is required for the transmutation. So a test linac is developed to accelerate high intensity beams. (N.K.)

  7. Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Jiang Jieqiong; Wang Minghuang; Chen Zhong; Qiu Yuefeng; Liu Jinchao; Bai Yunqing; Chen Hongli; Hu Yanglin

    2010-01-01

    Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

  8. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  9. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

  10. 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.

  11. Fission product chemistry in severe nuclear reactor accidents, specialists' meeting at JRC-Ispra, 15-17 January 1990

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-05-01

    A specialists' meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions). (author)

  12. Irradiation effects in fused quartz 'Suprasil' as a detector of fission fragments under high flux of reactor neutrons

    International Nuclear Information System (INIS)

    Moraes, O.M.G. de.

    1984-01-01

    A systematic study about the registration characteristics of synthetic fused quartz 'Suprasil I' use as a detector of fission fragments under high flux of reactor neutrons and the effects of irradiation on it was performed. Fission fragments of 252 Cf, gamma radiation doses of of 60 Co up to 150 MGy, and integrated neutrons fluxes up to 10 20 n/cm 2 were used. A model to explain the effects on track registration and development characteristics of 'Suprasil I' irradiated on reactors were proposed, based on the obtained results for efficiency an for annealing. (C.G.C.) [pt

  13. Design of a Mechanical NaK Pump for Fission Space Power Systems

    Science.gov (United States)

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

    2010-01-01

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

  14. 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

  15. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    International Nuclear Information System (INIS)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-01-01

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  16. 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

  17. 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.

  18. The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

    International Nuclear Information System (INIS)

    Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

    2012-01-01

    In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

  19. Modelling and simulation the radioactive source-term of fission products in PWR type reactors

    International Nuclear Information System (INIS)

    Porfirio, Rogilson Nazare da Silva

    1996-01-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  20. Natural fission reactors from Gabon. Contribution to the study of the conditions of stability of a natural radioactive wastes storage site (2 Ga)

    International Nuclear Information System (INIS)

    Pourcelot, L.

    1997-01-01

    The natural fission reactors of Oklo consists of a core of uraninite (60%) with fission products, embedded in a pure clay matrix. Thus, the aim of geological, mineral, and geochemical studies of the Oklo Reactors is to assess the behaviour of fission products in an artificial waste depository. Previous studies have shown that Reactor Zone 10, located in the Oklo mine, represents an example for an exceptional confinement of fission products since 2 Ga. In reactor Zone 9, located in Oklo open pit, migrations are more important. Reactor ZOne 13 was influenced by a thermal event due to a doleritic intrusion, located some twenty meters far away, one Ga years after fission reaction operations. In this study,we characterized temperature and redox conditions of fluids by using stable isotopes of uraninites and clays. Moreover mineralogical and chemical characteristics were defined. (author)

  1. 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

  2. NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stevenson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsai, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating.

  3. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  4. Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

    International Nuclear Information System (INIS)

    Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

    2009-01-01

    Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

  5. Computation of fission product distribution in core and primary circuit of a high temperature reactor during normal operation

    International Nuclear Information System (INIS)

    Mattke, U.H.

    1991-08-01

    The fission product release during normal operation from the core of a high temperature reactor is well known to be very low. A HTR-Modul-reactor with a reduced power of 170 MW th is examined under the aspect whether the contamination with Cs-137 as most important nuclide will be so low that a helium turbine in the primary circuit is possible. The program SPTRAN is the tool for the computations and siumlations of fission product transport in HTRs. The program initially developed for computations of accident events has been enlarged for computing the fission product transport under the conditions of normal operation. The theoretical basis, the used programs and data basis are presented followed by the results of the computations. These results are explained and discussed; moreover the consequences and future possibilities of development are shown. (orig./HP) [de

  6. Activation and Radiation Damage Behaviour of Russian Structural Materials for Fusion Reactors in the Fission and Fusion Reactors

    International Nuclear Information System (INIS)

    Blokhin, A.; Demin, N.; Chernov, V.; Leonteva-Smirnova, M.; Potapenko, M.

    2006-01-01

    Various structural low (reduced) activated materials have been proposed as a candidate for the first walls-blankets of fusion reactors. One of the main problems connected with using these materials - to minimise the production of long-lived radionuclides from nuclear transmutations and to provide with good technological and functional properties. The selection of materials and their metallurgical and fabrication technologies for fusion reactor components is influenced by this factor. Accurate prediction of induced radioactivity is necessary for the development of the fusion reactor materials. Low activated V-Ti-Cr alloys and reduced activated ferritic-martensitic steels are a leading candidate material for fusion first wall and blanket applications. At the present time a range of compositions and an impurity level are still being investigated to better understand the sensitive of various functional and activation properties to small compositional variations and impurity level. For the two types of materials mentioned above (V-Ti-Cr alloys and 9-12 % Cr f/m steels) and manufactured in Russia (Russia technologies) the analysis of induced activity, hydrogen and helium-production as well as the accumulation of such elements as C, N, O, P, S, Zn and Sn as a function of irradiation time was performed. Materials '' were irradiated '' by fission (BN-600, BOR-60) and fusion (Russian DEMO-C Reactor Project) typical neutron spectra with neutron fluency up to 10 22 n/cm 2 and the cooling time up to 1000 years. The calculations of the transmutation of elements and the induced radioactivity were carried out using the FISPACT inventory code, and the different activation cross-section libraries like the ACDAM, FENDL-2/A and the decay data library FENDL-2/D. It was shown that the level of impurities controls a long-term behaviour of induced activity and contact dose rate for materials. From this analysis the concentration limits of impurities were obtained. The generation of gas

  7. Development and optimization of neutron measurement methods by fission chamber on experimental reactors - management, treatment and reduction of uncertainties

    International Nuclear Information System (INIS)

    Blanc-De-Lanaute, N.

    2012-01-01

    The main objectives of this research thesis are the management and reduction of uncertainties associated with measurements performed by means of a fission-chamber type sensor. The author first recalls the role of experimental reactors in nuclear research, presents the various sensors used in nuclear detection (photographic film, scintillation sensor, gas ionization sensor, semiconducting sensor, other types of radiation sensors), and more particularly addresses neutron detection (activation sensor, gas filling sensor). In a second part, the author gives an overview of the state of the art of neutron measurement by fission chamber in a mock-up reactor (signal formation, processing and post-processing, associated measurements and uncertainties, return on experience of measurements by fission chamber on Masurca and Minerve research reactors). In a third part, he reports the optimization of two intrinsic parameters of this sensor: the thickness of fissile material deposit, and the pressure and nature of the filler gas. The fourth part addresses the improvement of measurement electronics and of post-processing methods which are used for result analysis. The fifth part deals with the optimization of spectrum index measurements by means of a fission chamber. The impact of each parameter is quantified. Results explain some inconsistencies noticed in measurements performed on the Minerve reactor in 2004, and allow the improvement of biases with computed values [fr

  8. Computer code and users' guide for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239b

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, R.A.; Smith, W.W.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The second volume describes the computer code and users' guide for the preliminary analysis of the system.

  9. Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

    Science.gov (United States)

    Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

    2017-10-24

    Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

  10. Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

    1993-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

  11. Fusion technology development: role of fusion facility upgrades and fission test reactors

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.

    1983-01-01

    The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. The authors show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

  12. Fusion technology development: role of fusion facility upgrades and fission test reactors

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Miller, L.G.; Longhurst, G.R.; Schmunk, R.E.

    1983-01-01

    The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. We will show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

  13. Local Fission Gas Release and Swelling in Water Reactor Fuel during Slow Power Transients

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg; Walker, C.T.; Ray, I.L.F.

    1985-01-01

    Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...... factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r/r0 ~ 0.75. Release tunnels were fully developed at r/r0 ~ 0.55 with the result that gas release was 60–70% at this position....

  14. Continuous radiochemical analysis of fission products in a nuclear reactor water coolant

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Zakharov, L.K.; Leont'ev, G.G.; Mel'nikov, V.A.; Orlenkov, I.S.; Slutskij, G.K.

    1975-01-01

    Method for continuous radiochemical analysis of I, Cs, Ba, Sr and Ce isotopes in a reactor water heat-transfer agent was developed. A continuous two-dimensional chromatographic process of complex mixtures separation of substances proved to be feasible on several parallel sorbent layers, which moved at constant velocities and separated by stationary intermediate collectors. Tests on model solutions containing I, Ce, Cs and Ba isotopes and on heat-carrier samples showed quantitative separation of elements. The results were indicative of a basic possibility of using multisorbent chromatographs for continuous control of multicomponent mixtures, particularly for control of radioactive fission product compositions in water heat-transfer agents in nuclear power plants. A diagram is shown for a two-dimensional chromatographic separation of a multicomponent mixture. Also shown is a flow chart of an installation for continuous control of iodine and cesium isotope activities

  15. Overview of Fusion-Fission Hybrid Reactor Design Study in China

    International Nuclear Information System (INIS)

    Huang Jinhua; Feng Kaiming; Deng Baiquan; Deng, P.Zh.; Zhang Guoshu; Hu Gang; He Kaihui; Wu Yican; Qiu Lijian; Huang Qunying; Xiao Bingjia; Liu Xiaoping; Chen Yixue; Kong, M.H.

    2002-01-01

    The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented

  16. Dosimetry of fission neutrons in a 1-W reactor, UTR-KINKI

    CERN Document Server

    Endo, S; Yoshitake, Y

    2002-01-01

    The energy spectrum of fission neutrons in the biological irradiation field of the Kinki University reactor, UTR-KINKI, has been determined by a multi-foil activation analysis coupled with artificial neural network techniques and a Au-foil activation method. The mean neutron energy was estimated to be 1.26+-0.05 MeV from the experimentally determined spectrum. Based on this energy value and other information, the neutron dose rate was estimated to be 19.7+-1.4 cGy/hr. Since this dose rate agrees with that measured by a pair of ionizing chambers (21.4 cGy/hr), we conclude that the mean neutron energy could be estimated with reasonable accuracy in the irradiation field of UTR-KINKI. (author)

  17. Method and apparatus for producing ultralowmass fissionable deposits for reactor neutron dosimetry by recoil ion-implantation

    International Nuclear Information System (INIS)

    Ruddy, F.H.

    1988-01-01

    A method for producing a fissionable deposit of selectively ultralow mass for neutron dosimetry is described comprising the steps of: (a) spacing in opposing relation a substrate and an alpha-emitting parent source which decays to implant into the substrate of fissionable daughter ejected from the parent source as a result of the decay; and (b) holding the opposing relation for a period of time until the parent source decays to form a corresponding mass of isotopically pure fissionable daughter uniformly on the substrate

  18. Cavity Ring-Down Spectroscopy for Gaseous Fission Products Trace Measurements in Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Jacquet, P.; Pailloux, A.; Doizi, D.; Aoust, G.; Jeannot, J.-P.

    2013-06-01

    Safety and availability are key issues of the generation IV reactors. Hence, the three radionuclide confinement barriers, including fuel cladding, must stay tight during the reactor operation. During the primary gaseous failure, fission products xenon and krypton are released. Their fast and sensitive detection guarantees the first confinement barrier tightness. In the frame of the French ASTRID project, an optical spectroscopy technique - Cavity Ring Down Spectroscopy (CRDS) - is investigated for the gaseous fission products measurement. A dedicated CRDS set-up is needed to detect the rare gases with a commercial laser. Indeed, the CRDS is coupled to a glow discharge plasma, which generates a population of metastable atoms. The xenon plasma conditions are optimized to 110 Pa and 1.3 W (3 mA). The production efficiency of metastable Xe is then 0.8 %, stable within 0.5% during hours. The metastable number density is proportional to the xenon over argon molar fraction. The spectroscopic parameters of the strong 823.16 nm xenon transition are calculated and/or measured in order to optimize the fit of the experimental spectra and make a quantitative measurement of the metastable xenon. The CRDS is coupled to the discharge cell. The laser intensity inside the cavity is limited by the optical saturation process, resulting from the strong optical pumping of the metastable state. The resulting weak CRDS signal requires a fast and very sensitive photodetector. A 600 ppt xenon molar fraction was measured by CRDS. With the present set-up, the detection limits are estimated from the baseline noise to approximately 20 ppt for each even isotope, 60 ppt for the 131 Xe and 55 ppt for the 129 Xe. This sensitivity matches the specifications required for gaseous leak measurement; approximately 100 ppt for 133 Xe (4 GBq/m 3 ) and 10 ppb for stable isotopes. The odd isotopes are selectively measured, whereas the even isotopes overlap, a spectroscopic feature that applies for stable or

  19. Power deposition distribution in liquid lead cooled fission reactors and effects on the reactor thermal behaviour

    International Nuclear Information System (INIS)

    Cevolani, S.; Nava, E.; Burn, K. W.

    2001-01-01

    In the framework of an ADS study (Accelerator Driven System, a reactor cooled by a lead bismuth alloy) the distribution of the deposited energy between the fuel, coolant and structural materials was evaluated by means of Monte Carlo calculations. The energy deposition in the coolant turned out to be about four percent of the total deposited energy. In order to study this effect, further calculations were performed on water and sodium cooled reactors. Such an analysis showed, for both coolant materials, a much lower heat deposition, about one percent. Based on such results, a thermohydraulic analysis was performed in order to verify the effect of this phenomenon on the fuel assembly temperature distribution. The main effect of a significant fraction of energy deposition in the coolant is concerned with the decrease of the fuel pellet temperature. As a consequence, taking into account this effect allows to increase the possibilities of optimization at the disposal of the designer [it

  20. 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.

  1. Comparison of 252Cf time correlated induced fission with AmLi induced fission on fresh MTR reserach reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jay Prakash [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-01

    The objectives of this project are to calibrate the Advanced Experimental Fuel Counter (AEFC), benchmark MCNP simulations using experimental results, investigate the effects of change in fuel assembly geometry, and finally to show the boost in doubles count rates with 252Cf active soruces due to the time correlated induced fission (TCIF) effect.

  2. Radiochemical studies on fission

    Energy Technology Data Exchange (ETDEWEB)

    None

    1973-07-01

    Research progress is reported on nuclear chemistry; topics considered include: recoil range and kinetic energy distribution in the thermal neutron ftssion of /sup 245/Cm; mass distribution and recoil range measurements in the reactor neutron-induced fission of /sup 232/U; fission yields in the thermal neutron fission of /sup 241/PU highly asymmetric binary fission of uranium induced by reactor neutrons; and nuclear charge distribution in low energy fission. ( DHM)

  3. Feasibility study of a small, thorium-based fission power system for space and terrestrial applications

    Science.gov (United States)

    Worrall, Michael Jason

    One of the current challenges facing space exploration is the creation of a power source capable of providing useful energy for the entire duration of a mission. Historically, radioisotope batteries have been used to provide load power, but this conventional system may not be capable of sustaining continuous power for longer duration missions. To remedy this, many forays into nuclear powered spacecraft have been investigated, but no robust system for long-term power generation has been found. In this study, a novel spin on the traditional fission power system that represents a potential optimum solution is presented. By utilizing mature High Temperature Gas Reactor (HTGR) technology in conjunction with the capabilities of the thorium fuel cycle, we have created a light-weight, long-term power source capable of a continuous electric power output of up to 70kW for over 15 years. This system relies upon a combination of fissile, highly-enriched uranium dioxide and fertile thorium carbide Tri-Structural Isotropic (TRISO) fuel particles embedded in a hexagonal beryllium oxide matrix. As the primary fissile material is consumed, the fertile material breeds new fissile material leading to more steady fuel loading over the lifetime of the core. Reactor control is achieved through an innovative approach to the conventional boron carbide neutron absorber by utilizing sections of borated aluminum placed in rotating control drums within the reflector. Borated aluminum allows for much smaller boron concentrations, thus eliminating the potential for 10B(n,alpha)6Li heating issues that are common in boron carbide systems. A wide range of other reactivity control systems are also investigated, such as a radially-split rotating reflector. Lastly, an extension of the design to a terrestrial based system is investigated. In this system, uranium enrichment is dropped to 20 percent in order to meet current regulations, a solid uranium-zirconium hydride fissile driver replaces the

  4. Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

    International Nuclear Information System (INIS)

    Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

    2000-01-01

    Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

  5. Euratom innovation in nuclear fission: Community research in reactor systems and fuel cycles

    International Nuclear Information System (INIS)

    Goethem, G. van; Hugon, M.; Bhatnagar, V.; Manolatos, P.; Deffrennes, M.

    2007-01-01

    The following questions are naturally at the heart of the current Euratom research and training framework programme:(1)What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2)What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy, but also more generally as is depicted in the following figure. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle' in above figure) respond to the following long-term criteria: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. Research and innovation in nuclear fission technology has broad and extended geographical, disciplinary and time horizons:- the community involved extends to all 25 EU Member States and beyond; - the research assembles a large variety of scientific disciplines; - three generations of nuclear power technologies (called II, III and IV) are involved, with the timescales extending from now to around the year 2040. To each of these three generations, a couple of challenges are associated (six in total):- Generation II (1970-2000, today): security of supply+environmental compatibility; - Generation III (around 2010): enhanced safety and competitiveness (economics); - Generation IV (around 2040): cogeneration of heat and power, and full recycling. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is

  6. Neutronic performance of a fusion-fission hybrid reactor designed for fuel enrichment for LWRs

    International Nuclear Information System (INIS)

    Yapici, H.; Baltacioglu, E.

    1997-01-01

    In this study, the breeding performance of a fission hybrid reactor was analyzed to provide fissile fuel for Light Water Reactors (LWR) as an alternative to the current methods of gas diffusion and gas centrifuge. LWR fuel rods containing UO 2 or ThO 2 fertile material were located in the fuel zone of the blanket and helium gas or Flibe (Li 2 BeF 4 ) fluid was used as coolant. As a result of the analysis, according to fusion driver (D,T and D,D) and the type of coolant the enrichment of 3%-4% were achieved for operation periods of 12 and 36 months in case of fuel rods containing UO 2 , respectively and for operation periods of 18 and 48 months in case of fuel rods containing ThO 2 , respectively. Depending on the type of fusion driver, coolant and fertile fuel, varying enrichments of between 3% and 8.9% were achieved during operation period of four years

  7. Upper-bound fission product release assessment for large break LOCA in CANFLEX bundle reactor core

    International Nuclear Information System (INIS)

    Oh, Duk Ju; Lee, Kang Moon

    1996-07-01

    Quarter-core gap inventory assessment for CANDU-6 reactor core loaded with CANFLEX fuel bundles has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The quarter-core gap inventory for the CANFLEX bundle core is 5 - 10 times lower than that for the standard bundle core, depending on the half-life of the isotope. The lower gap inventory of the CANFLEX bundle core is attributed to the lower linear power of the CANFLEX bundle compared with the standard bundle. However, the whole core total inventories for both the CANFLEX and standard bundle cores are nearly the same. The 6 - 8 times lower upper-bound fission product releases of the CANFLEX bundle core for large break LOCA than those of the standard bundle core imply that the loading of 24 natural uranium CANFLEX bundles would improve the predicted consequences of the postulated accident described in the Wolsung 2 safety report. 2 tabs., 6 figs., 3 refs. (Author)

  8. Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2002-07-01

    Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10 -2 Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10 -3 Tdk/dT. (author)

  9. 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

  10. 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)

  11. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

    Behrens, J.W.; Carlson, A.D.

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, ''Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately

  12. The unique safety challenges of space reactor systems

    International Nuclear Information System (INIS)

    Lanes, S.J.; Marshall, A.C.

    1991-01-01

    Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

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

    International Nuclear Information System (INIS)

    Marshall, A.C.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1995-12-31

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

  15. Transmutation of Tc-99 and I-129 in fission reactors. A calculational study

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1995-03-01

    The HWR is a better candidate for large-scale transmutation of long-lived fission products. When target pins containing either Tc-99 or I-129 are positioned in the centre of each fuel bundle of a 935 MW e CANDU reactor, the transmutation half lives are 44 and 20 years, respectively, and the gross transmutation rates 60 and 48 kg/a. The positive coolant void coefficient is reduced in both cases with about 30%. When Tc-99 target pins are positioned in the moderator between the fuel bundles, the transmutation half life becomes 25 years and the gross transmutation rate 106 kg/a. This means that one HWR can serve four PWRs with equal power. The fast reactor seems most promising. When Tc-99 target pins are irradiated in moderated subassemblies in the inner core of Superphenix (∼1240 MW e ), a transmutation half life of 15 years is obtained with a gross transmutation rate of 122 kg/a. These values become 18 years and 101 kg/a when non-moderated subassemblies are used for the irradiation. This implies that one fast reactor can serve four to five PWRs with equal power. The PWR seems not very effective for transmutation of Tc-99. Large inventories are needed to obtain a Tc-99 transmutation rate equal to the production rate (18 kg/a for a 900 MW e PWR). When all guide tubes of an UO 2 fuelled PWR are filled with Tc-99 with density of 5 g cm -3 , the transmutation half life is 39 years and the gross transmutation rate 64 kg/a. (orig./GL)

  16. Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

    International Nuclear Information System (INIS)

    Okrent, D.

    1997-01-01

    This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, with the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident

  17. Miniature fission chambers calibration in pulse mode: interlaboratory comparison at the. SCK·CEN BR1 and CEA CALIBAN reactors

    International Nuclear Information System (INIS)

    Lamirand, V.; Geslot, B.; Gregoire, G.; Garnier, D.; Breaud, S.; Mellier, F.; Di-Salvo, J.; Destouches, C.; Blaise, P.; Wagemans, J.; Borms, L.; Malambu, E.; Casoli, P.; Jacquet, X.; Rousseau, G.; Sauvecane, P.

    2013-06-01

    Miniature fission chambers are suited tools for instrumenting experimental reactors, allowing online and in-core neutron measurements of quantities such as fission rates or reactor power. A new set of such detectors was produced by CEA to be used during the next experimental program at the EOLE facility starting in 2013. Some of these detectors will be employed in pulse mode for absolute measurements, thus requiring calibration. The calibration factor is expressed in mass units and thus called 'effective mass'. A calibration campaign was conducted in December 2012 at the SCK.CEN BR1 facility within the framework of the scientific cooperation VEP (VENUS-EOLE-PROTEUS) between SCK.CEN, CEA and PSI. Two actions were conducted in order to improve the calibration method. First a new characterisation of the thermal flux cavity and the MARK3 neutron flux conversion device performed by SCK.CEN allowed using calculated effective cross sections for determining detectors effective masses. Dosimetry irradiations were performed in situ in order to determine the neutron flux level and provide link to the metrological standard. Secondly two fission chambers were also calibrated at the CEA CALIBAN reactor (fast neutron spectrum), using the same method so that the results can be compared with the results obtained at the SCK.CEN. In this paper the calibration method and recent improvements on uncertainty reduction are presented. The results and uncertainties obtained in the two reactors CALIBAN and BR1 are compared and discussed. (authors)

  18. Organic free radicals and micropores in solid graphitic carbonaceous matter at the Oklo natural fission reactors, Gabon

    International Nuclear Information System (INIS)

    Rigali, M.J.; Nagy, B.

    1997-01-01

    The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radio nuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radio nuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radio nuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 ± 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modem humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. 72 refs., 7 figs., 1 tab

  19. Neutronic analysis for the fission Mo99 production by irradiation of leu targets in TRIGA 14 MW reactor

    International Nuclear Information System (INIS)

    Dulugeac, S. D.; Mladin, M.; Budriman, A. G.

    2013-01-01

    Molybdenum production can be a solution for the future in the utilization of the Romanian TRIGA, taking into account the international market supply needs. Generally, two different techniques are available for Mo 99 production for use in medical Tc 99 generation.The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in Mo 98 . In a second process, Mo 99 is obtained as a result of the neutron induced fission of U 235 according to U 235 (n,f) Mo 99 . The objectives of the paper are related to Mo 99 production as a result of fission. Neutron physics parameters are determined and presented, such as: thermal flux axial distribution for the critical reactor at 10 MW inside the irradiation location; reactivity introduced by three Uranium foil containers; neutron fluxes and fission rates in the Uranium foils; released and deposited power in the Uranium foils; Mo 99 activity in the Uranium foils. (authors)

  20. 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

  1. Experiments to determine the rate of beta energy release following fission of Pu239 andU235 in a fast reactor

    International Nuclear Information System (INIS)

    Murphy, M.F.; Taylor, W.H.; Sweet, D.W.; March, M.R.

    1979-02-01

    Measurements have been made of the rate of beta energy release from Pu239 and U235 fission fragments over a period of 107 seconds following a 105 second irradiation in the zero-power fast reactor Zebra. Results are compared with predictions using the UKFPDD-1 decay data file and two different sets of fission product yield data. (author)

  2. 239Pu prompt fission neutron spectra impact on a set of criticality and experimental reactor benchmarks

    International Nuclear Information System (INIS)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

    2014-01-01

    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239 Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239 Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  3. A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho; Pyeon, Cheol Ho

    2015-01-01

    In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r g , E g , t g ) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the neutron sources

  4. A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Pyeon, Cheol Ho [Kyoto University, Osaka (Japan)

    2015-10-15

    In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r{sub g}, E{sub g}, t{sub g}) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the

  5. 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

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

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

  8. Mathematical model for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239a

    Energy Technology Data Exchange (ETDEWEB)

    Grey, J.; Chow, S.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. Such a concept could be particularly useful for missions which require both relatively high acceleration (e.g., for planetocentric maneuvers) and high performance at low acceleration (e.g., on heliocentric trajectories or for trajectory shaping). The first volume develops the mathematical model of the system.

  9. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    International Nuclear Information System (INIS)

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs

  10. Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

    International Nuclear Information System (INIS)

    Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.

    1980-01-01

    Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

  11. Calculated model of radioactive fission and corrosion product accumulation and distribution in a fast reactor sodium coolant circuit

    International Nuclear Information System (INIS)

    Kizin, V.D.; Konyashov, V.V.

    1987-01-01

    A simple calculation procedure of radioactive products accumulation and distribution in a primary circuit has been developed on the basis of experimental investigations at the BOR-60 reactor. Common knowledge on the impurity products transfer at the liquid-solid and liquid-gas phase boundary is taken. Use is made of the typical in reactor physics relationships for the description of the products transition to the equipment surfaces, of fission products release, metal corrosion and others. Satisfactory agreement of the calculation data with the experimental ones has been obtained. (orig.)

  12. Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

    International Nuclear Information System (INIS)

    Ku, L.P.; Price, W.G. Jr.

    1977-08-01

    The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

  13. Water reactor fuel behaviour and fission products release in off-normal and accident conditions

    International Nuclear Information System (INIS)

    1987-09-01

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

  14. Fission-product chemistry in severe reactor accidents: Review of relevant integral experiments

    International Nuclear Information System (INIS)

    Nichols, A.L.; Hueber, C.

    1992-01-01

    The attenuation of the radioactive fission-product emission from a severe reactor accident will depend on a combination of chemical, physical and thermal-hydraulic effects. Chemical species stabilised under the prevailing conditions will determine the extent of aerosol formation and any subsequent interaction, so defining the magnitude and physical forms of the eventual release into the environment. While several important integral tests have taken place in recent years, these experiments have tended to focus on the generation of mass-balance and aerosol-related data to test and validate materials-transport codes rather than study the impact of important chemical phenomena. This emphasis on thermal hydraulics, fuel behaviour and aerosol properties has occurred in many test (e.g. PBF, DEMONA, Marviken-V, LACE and ACE). Nevertheless, the generation and reaction of the chemical species in all of these programmes determined the transport properties of the resulting vapours and aerosols. Chemical effects have been studied in measurements somewhat subsidiary to the main aims of the tests. This work has been reviewed in detail with respect to Marviken-V, LACE, ACE and Falcon. Specific issues remain to be addressed, and these are discussed in terms of the proposed Phebus-FB programme. (author). 58 refs, 9 figs, 1 tab

  15. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    Science.gov (United States)

    Capelli, E.; Beneš, O.; Konings, R. J. M.

    2018-04-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all the common-ion binary and ternary sub-systems of the LiF-ThF4-CsF-LiI-ThI4-CsI system has been performed and the optimized parameters are presented in this work. New equilibrium data have been measured using Differential Scanning Calorimetry and were used to assess the reciprocal ternary systems and confirm the extrapolated phase diagrams. The developed database significantly contributes to the understanding of the behaviour of cesium and iodine in the MSR, which strongly depends on their concentration and chemical form. Cesium bonded with fluorine is well retained in the fuel mixture while in the form of CsI the solubility of these elements is very limited. Finally, the influence of CsI and CsF on the physico-chemical properties of the fuel mixture was calculated as function of composition.

  16. Yields of some fragments on 235U, 238U and 239Pu fission due to the neutrons of the SBR-1 reactors

    International Nuclear Information System (INIS)

    Yurova, L.N.; Bushuev, A.V.; Ozerkov, V.N.; Chachin, V.V.; Zvonarev, A.V.; Liforov, Yu.G.; Koleganov, Yu.V.; Miller, V.V.; Gorbatyuk, O.V.

    1979-01-01

    Determined are the values of the yields of fission fragments in spectrum close to that of the neutron fission using the data on yields at fission by thermal neutrons. The relation between the activities of fragments in samples irradiated in the BR-1 center and in the thermal colomn of the same reactor was measured with the help of the Ge(Li). The relative rate of fissions in uranium and plutonium samples in the center or in thermal colomn were measured by track detectors. The comparison of the yields obtained and the data of other authors is being made

  17. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

    Behrens, J.W.; Carlson, A.D.

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, ''Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately

  18. Survey of Materials for Fusion Fission Hybrid Reactors Vol 1 Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Joseph Collin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Chemistry Materials and Life Sciences Directorate

    2007-07-03

    Materials for fusion-fission hybrid reactors fall into several broad categories, including fuels, blanket and coolant materials, cladding, structural materials, shielding, and in the specific case of inertial-confinement fusion systems, laser and optical materials. This report surveys materials in all categories of materials except for those required for lasers and optics. Preferred collants include two molten salt mixtures known as FLIBE (Li2BeF4) and FLINABE (LiNaBeF4). In the case of homogenous liquid fuels, UF4 can be dissolved in these molten salt mixtures. The transmutation of lithium in this coolant produces very corrosive hydrofluoric acid species (HF and TF), which can rapidly degrade structural materials. Broad ranges of high-melting radiation-tolerant structural material have been proposed for fusion-fission reactor structures. These include a wide variety of steels and refractory alloys. Ferritic steels with oxide-dispersion strengthening and graphite have been given particular attention. Refractory metals are found in Groups IVB and VB of the periodic table, and include Nb, Ta, Cr, Mo, and W, as serve as the basis of refractory alloys. Stable high-melting composites and amorphous metals may also be useful. Since amorphous metals have no lattice structure, neutron bombardment cannot dislodge atoms from lattice sites, and the materials would be immune from this specific mode of degradation. The free energy of formation of fluorides of the alloying elements found in steels and refractory alloys can be used to determine the relative stability of these materials in molten salts. The reduction of lithium transmutation products (H+ and T+) drives the electrochemical corrosion process, and liberates aggressive fluoride ions that pair with ions formed from dissolved structural materials. Corrosion can be suppressed through the use of metallic Be and Li, though the molten salt becomes laden with colloidal suspensions of Be and Li corrosion

  19. The role of nuclear reactors in space exploration and development

    International Nuclear Information System (INIS)

    Lipinski, R.J.

    2000-01-01

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

  20. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    International Nuclear Information System (INIS)

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

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

    Science.gov (United States)

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

    2017-08-01

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

  2. Reconstruction of intra-bundle fission density profile during a postulated LOCA in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, D. [Oak Ridge National Laboratory (United States); Rahnema, F. [Georgia Inst. of Technology (United States); Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Inst. of Technology, Atlanta, GA 30332-0405 (United States); Serghiuta, D. [Canadian Nuclear Safety Commission (Canada); Sarsour, H.; Turinsky, P. J. [North Carolina State Univ. (United States); Stamm' ler, R. [Studsvik Scandpower AS (Norway)

    2006-07-01

    In this paper, results related to the reconstruction of intra-bundle fission density profile for a 37-pin CANDU-6 bundle with the highest enthalpy deposition during a postulated large LOCA stagnation break in a Bruce B core are presented. Bruce B is a nuclear power plant in Kincardine, Ontario (Canada)), on the shores of Lake Huron with 4 CANDU reactors that are rated at about 750 MWe. The reconstruction of the fuel pin fission densities is based on steady-state, three-dimensional simulations with the Monte Carlo code MCNP for a subset of 27 out of 69 time steps during the first two seconds of the power pulse predicted for the fuel bundle at core location V13/8. Two-group cross section data libraries are generated for MCNP at each time step by the lattice depletion neutron transport code HELIOS-1.7. To include the effect of the surrounding core environment, the calculations are performed with time-dependent albedo boundary conditions inferred from a full core simulation of the transient by the nodal diffusion code NESTLE with HELIOS homogenized cross-sections. It is found that the local peaking factor (LPF) in the outer ring varies during the transient, but never exceeds its value before the transient. Inclusion of the core environment increases the LPF in the outer ring. For the analyzed case, the increase is 0.72% with a relative error of 0.01% for the LPF before the transient and 0.55% (with a relative error of 0.01%) for the maximum average LPF during the transient. The latter is based on only four selected transient time points. Note that the immediate environment of the 'hot bundle' does not contain any reactivity devices or other perturbing factors. As a result, the increases observed in the LPF in the outer ring may not be representative of the situations in which 'other' core environment perturbing factors are present. To determine the effect of these factors on the LPF, further analyses of a bundle in the proximity of control devices

  3. 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

  4. Growth of optical transmission loss at 850 nm in silica core optical fibers during fission reactor irradiation

    International Nuclear Information System (INIS)

    Shikama, T.; Narui, M.; Sagawa, T.

    1998-01-01

    Pure, OH-doped and F-doped silica core optical fibers were irradiated in a fission reactor at 400±10 K using an electric heater at a reactor power greater than 10 MW (20% of the full power). The temperature was not controlled well at the early stage of the reactor startup, when the temperature was about 320-340 K. The optical fibers were irradiated with a fast neutron (E>1 MeV) flux of 3.2 x 10 17 n/cm 2 s and a gamma dose rate of 3 x 10 3 Gy/s for 527 h. Optical transmission loss at 850 nm was measured in situ during irradiation. A prompt increase in optical transmission loss was observed as irradiation started, which was probably due to dynamic irradiation effects caused by short-lived and transient defects and is probably recoverable when irradiation ceases. After the prompt increase in optical transmission loss, a so-called radiation hardening was observed in fibers containing OH. Radiation hardening was also observed in 900 ppm OH-doped fiber at the second startup. The optical transmission loss increased linearly with irradiation dose, denoted as the accumulated loss, which we believe is due to irradiation-induced long-lived defects. Accumulated loss dominates radiation-induced optical transmission loss in a fission reactor irradiation. (orig.)

  5. The Oklo natural nuclear reactors: neutron parameters, age and duration of the reactions, uranium and fission products migrations

    International Nuclear Information System (INIS)

    Ruffenach, J.-C.

    1979-09-01

    Mass spectrometry and isotopic dilution technique are used in order to carry out, on various samples from the fossil nuclear reactors at Oklo, Gabon, isotopic and chemical analyses of some particular elements involved in the nuclear reactions: uranium, lead, bismuth, thorium, rare gases (krypton, xenon), rare earths (neodymium, samarium, europium, gadolinium, dysprosium), ruthenium and palladium. Interpretations of these analyses lead to the determination of many neutron parameters such as the neutron fluence received by the samples, the spectrum index, the conversion coefficient, and also the percentages of fissions due to uranium-238 and plutonium-239 and the total number of fissions relative to uranium. All these results make it possible to determine the age of the nuclear reactions by measuring the amounts of fission rare earths formed, i.e. 1.97 billion years. This study brings some informations to the general problem of radioactive wastes storage in deep geological formations, the storage of uranium, plutonium and many fission products having been carried out naturally, and for about two billion years [fr

  6. 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

  7. 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...

  8. Power-balance analysis of muon-catalyzed fusion-fission hybrid reactor systems

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1985-01-01

    A power-balance model of a muon-catalyzed fusion system in the context of a fission-fuel factory is developed and exercised to predict the required physics performance of systems competitive with either pure muon-catalyzed fusion systems or thermonuclear fusion-fission fuel factory hybrid systems

  9. 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.)

  10. Two billion year old natural analogs for nuclear waste disposal: the natural nuclear fission reactors in Gabon (Africa)

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.

    2002-01-01

    Two billion years ago, the increase of oxygen in atmosphere and the high 235 U/ 238 U uranium ratio (> 3%) made possible the occurrence of natural nuclear reactors on Earth. These reactors are considered to be a good natural analogue for nuclear waste disposal. Their preservation during such a long period of time is mainly due to the geological stability of the site, the occurrence of clays surrounding the reactors and acting as an impermeable shield, and the occurrence of organic matter that maintained the environment in reducing conditions, favourable for the stability of uraninite. Hydrogeochemical studies and modelling have shown the complexity of the geochemical system at Oklo and Bangombe (Gabon) and the lack of precise data about uranium and fission products retention and migration mechanisms in geological environments. (author)

  11. Fission reactor based epithermal neutron irradiation facilities for routine clinical application in BNCT-Hatanaka memorial lecture

    International Nuclear Information System (INIS)

    Harling, Otto K.

    2009-01-01

    Based on experience gained in the recent clinical studies at MIT/Harvard, the desirable characteristics of epithermal neutron irradiation facilities for eventual routine clinical BNCT are suggested. A discussion of two approaches to using fission reactors for epithermal neutron BNCT is provided. This is followed by specific suggestions for the performance and features needed for high throughput clinical BNCT. An example of a current state-of-the-art, reactor based facility, suited for routine clinical use is discussed. Some comments are provided on the current status of reactor versus accelerator based epithermal neutron sources for BNCT. This paper concludes with a summary and a few personal observations on BNCT by the author.

  12. 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)

  13. Fission product data for thermal reactors. Final report. Part 2. Users manual for EPRI-CINDER code and data

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.; Stamatelatos, M.G.

    1976-12-01

    A four-group fission-product absorption chain library using ENDF/B-IV decay data and cross sections processed with a typical light water reactor spectrum for a modified version of the original CINDER code has been developed as described in Part 1. CINDER is a general point-depletion and fission product code based on an analytical solution of the equations describing nuclides coupled in any linear sequence of radioactive decays and neutron absorptions. The basic code has been in wide use for a number of years. Previously, the user was required to specify all physical data. This report describes the chain library in detail and a modified version of the basic CINDER code (EPRI-CINDER) that is still compatible with existing libraries

  14. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  15. 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

  16. Fission products collecting devices

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1979-01-01

    Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

  17. Neutral space analysis for a Boolean network model of the fission yeast cell cycle network

    Directory of Open Access Journals (Sweden)

    Gonzalo A Ruz

    2014-01-01

    Full Text Available BACKGROUND: Interactions between genes and their products give rise to complex circuits known as gene regulatory networks (GRN that enable cells to process information and respond to external stimuli. Several important processes for life, depend of an accurate and context-specific regulation of gene expression, such as the cell cycle, which can be analyzed through its GRN, where deregulation can lead to cancer in animals or a directed regulation could be applied for biotechnological processes using yeast. An approach to study the robustness of GRN is through the neutral space. In this paper, we explore the neutral space of a Schizosaccharomyces pombe (fission yeast cell cycle network through an evolution strategy to generate a neutral graph, composed of Boolean regulatory networks that share the same state sequences of the fission yeast cell cycle. RESULTS: Through simulations it was found that in the generated neutral graph, the functional networks that are not in the wildtype connected component have in general a Hamming distance more than 3 with the wildtype, and more than 10 between the other disconnected functional networks. Significant differences were found between the functional networks in the connected component of the wildtype network and the rest of the network, not only at a topological level, but also at the state space level, where significant differences in the distribution of the basin of attraction for the G1 fixed point was found for deterministic updating schemes. CONCLUSIONS: In general, functional networks in the wildtype network connected component, can mutate up to no more than 3 times, then they reach a point of no return where the networks leave the connected component of the wildtype. The proposed method to construct a neutral graph is general and can be used to explore the neutral space of other biologically interesting networks, and also formulate new biological hypotheses studying the functional networks in the

  18. Nuclear data for fission reactor core design and safety analysis: Requirements and status of accuracy of nuclear data

    International Nuclear Information System (INIS)

    Rowlands, J.L.

    1984-01-01

    The types of nuclear data required for fission reactor design and safety analysis, and the ways in which the data are represented and approximated for use in reactor calculations, are summarised first. The relative importance of different items of nuclear data in the prediction of reactor parameters is described and ways of investigating the accuracy of these data by evaluating related integral measurements are discussed. The use of sensitivity analysis, together with estimates of the uncertainties in nuclear data and relevant integral measurements, in assessing the accuracy of prediction of reactor parameters is described. The inverse procedure for deciding nuclear data requirements from the target accuracies for prediction of reactor parameters follows on from this. The need for assessments of the uncertainties in nuclear data evaluations and the form of the uncertainty information is discussed. The status of the accuracies of predictions and nuclear data requirements are then summarised. The reactor parameters considered include: (a) Criticality conditions, conversion and burn-up effects. (b) Energy production and deposition, decay heating, irradiation damage, dosimetry and induced radioactivity. (c) Kinetics characteristics and control, including temperature, power and coolant density coefficients, delayed neutrons and control absorbers. (author)

  19. Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

    2000-01-01

    This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

  20. Feasibility study of chabazite absorber tube utilization in online absorption of released gaseous fission products and substitution of burnable absorber rods with chabazite absorber tubes in VVER-1000 reactor series

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • Chabazite tubes are used for online removal of the released gaseous fission products. • The feasibility of using chabazite tubes instead of burnable absorber rods was studied. • A computational cycle was designed using the WIMSD5-B, CITATION-LDI2 and WERL codes. • In modeling fission gas release, the Weisman, Booth, Mason and T.S. models were used. • By this method, it is possible to increase cycle length and enhance heat transfer. - Abstract: As gaseous fission products, e.g. xenon and krypton have adverse effects such as reducing the rate of heat transfer in fuel rods and adding negative reactivity to the reactor core, the present manuscript was dedicated to development of a novel method for improving these defects. In the proposed method, chabazite absorber tubes were used for online removal of the released gaseous fission products from gaseous gap spaces. Moreover, in this research, feasibility of using chabazite absorber tubes instead of burnable absorber rods was examined. To perform the required modeling and calculations to successfully meet the mentioned objectives, a thermo-neutronic computational cycle was designed using the coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic calculations. In addition, in modeling the release process of gaseous fission products, the Weisman, Booth, Mason, and T.S. models were examined. It is worth mentioning that in this research, calculations and modeling procedures were based on the first cycle of Bushehr’s VVER-1000 reactor to study the feasibility of the proposed solution. The obtained results revealed that with application of the proposed method in this research, it is possible to increase cycle length, improve safety thresholds, and enhance heat transfer in the core of nuclear reactors.

  1. 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.

  2. On fission product retention in the core of the low powered high temperature reactor under accident conditions

    International Nuclear Information System (INIS)

    Bastek, H.

    1984-01-01

    In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

  3. Low-cost space fission power systems utilizing US and former Soviet Union experience and technology

    International Nuclear Information System (INIS)

    Wetch, J.R.; Britt, E.J.; Koester, J.K.; Gunther, N.; Ponomarev-Stepnoi, N.N.; Nikolaev, Y.V.; Nikitin, V.

    1997-01-01

    This report summarizes the author close-quote s approach to space power total economics. In the past 40 years of U.S. government sponsored space nuclear power developments, total economics has received only token consideration. In the real world, nuclear power has had limited acceptance where it provided the enabling capability i.e. isotopes for low power, long life, deep space missions, or reactor power for underwater nuclear submarines. It was also accepted where it was perceived to be more economic. Examples are nuclear reactor powered aircraft carriers, escort vessels and central station power stations. In any case, real and perceived public and environmental safety must always be included into the economic equation. copyright 1997 American Institute of Physics

  4. High temperature fission chambers. Fast breeder reactor research and development program

    International Nuclear Information System (INIS)

    Berlin, C.; Perrigueur, J.C.

    1984-04-01

    Development of a high temperature fission chamber and experimentations of measuring channels (detectors and electronic devices) in similar conditions as those of power plants: development of measuring channels (impulses and current) of the Super Phenix neutronic measures auxiliary system, development of a measuring channel with impulses for the surveillance system of the clad failures, based on integrated detectors, and development of a fission chamber for experimentations in similar conditions as in Superphenix [fr

  5. Tritium control and capture in salt-cooled fission and fusion reactors: Status, challenges, and path forward

    International Nuclear Information System (INIS)

    Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; Whyte, Dennis G.; Scarlat, Raluca

    2017-01-01

    Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The Fluoride-salt-cooled High-temperature Reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the base-line salts contain lithium where isotopically separated "7Li is proposed to minimize tritium production from neutron interactions with the salt. The Chinese Academy of Science plans to start operation of a 2-MWt molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in "6Li is proposed to maximize tritium generation the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700 °C liquid salt systems. We describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data is the primary constraint for designing efficient cost-effective methods of tritium control.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  12. 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.

  13. The story of fission reactors: from Chicago Pile to advanced energy systems

    International Nuclear Information System (INIS)

    Kannan, Umasankari

    2017-01-01

    Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

  14. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  15. Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

    International Nuclear Information System (INIS)

    Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

    2007-01-01

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

  16. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  17. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    Science.gov (United States)

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  18. Estimation, comparison, and evaluation of advanced fission power reactor generation costs

    International Nuclear Information System (INIS)

    Waddell, J.D.

    1977-01-01

    The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

  19. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release

    International Nuclear Information System (INIS)

    Zacharie, I.

    1997-01-01

    In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author)

  20. Shield materials recommended for space power nuclear reactors

    Science.gov (United States)

    Kaszubinski, L. J.

    1973-01-01

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

  1. Joint ICFRM-14 (14. international conference on fusion reactor materials) and IAEA satellite meeting on cross-cutting issues of structural materials for fusion and fission applications. PowerPoint presentations

    International Nuclear Information System (INIS)

    2009-01-01

    The Conference was devoted to the challenges in the development of new materials for advanced fission, fusion and hybrid reactors. The topics discussed include fuels and materials research under the high neutron fluence; post-irradiation examination; development of radiation resistant structural materials utilizing fission research reactors; core materials development for the advanced fuel cycle initiative; qualification of structural materials for fission and fusion reactor systems; application of charged particle accelerators for radiation resistance investigations of fission and fusion structural materials; microstructure evolution in structural materials under irradiation; ion beams and ion accelerators

  2. Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Massod; Bokhari, Ishtiaq Hussain; Mahmood, Tariq; Mahmood, Tayyab; Ahmad, Zahoor; Zaman, Qamar [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-02-15

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% {sup 235}U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.

  3. Energy released in fission

    International Nuclear Information System (INIS)

    James, M.F.

    1969-05-01

    The effective energy released in and following the fission of U-235, Pu-239 and Pu-241 by thermal neutrons, and of U-238 by fission spectrum neutrons, is discussed. The recommended values are: U-235 ... 192.9 ± 0.5 MeV/fission; U-238 ... 193.9 ± 0.8 MeV/fission; Pu-239 ... 198.5 ± 0.8 MeV/fission; Pu-241 ... 200.3 ± 0.8 MeV/fission. These values include all contributions except from antineutrinos and very long-lived fission products. The detailed contributions are discussed, and inconsistencies in the experimental data are pointed out. In Appendix A, the contribution to the total useful energy release in a reactor from reactions other than fission are discussed briefly, and in Appendix B there is a discussion of the variations in effective energy from fission with incident neutron energy. (author)

  4. 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.

  5. 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

  6. CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors

    International Nuclear Information System (INIS)

    England, T.R.; Gorrell, T.C.; Hightower, J.H.

    2001-01-01

    1 - Description of problem or function: CINDER is a four-group, one- point depletion and fission product program based on the evaluation of a general analytical solution of nuclides coupled in any linear sequence of radioactive decays and neutron absorptions in a specified neutron flux spectrum. The desired depletion and fission product chains and all physical data are specified by the problem originator. The program computes individual nuclide number densities, activities, nine energy-group disintegration rates, and macroscopic and barns/fission poisons at each time-step as well as selected summaries of these data. 2 - Method of solution: Time-dependent variations in nuclide cross sections and neutron fluxes are approximated by a user-specified sequential set of values which are considered constant during the duration of the user-specified associated time-increments. When a nuclide concentration is independent of the concentration of any of its progeny, it is possible to resolve the couplings so as to obtain nuclides fed by a single parent. These chains are referred to as linear. 3 - Restrictions on the complexity of the problem: The program is limited to 500 total nuclides formed in up to 240 chains of 20 or fewer nuclides each. Up to 10 nuclides may act as fission product sources, contributing to power, and as many as 99 time-steps of arbitrary length are permitted. All stable nuclides must have a cross section if zero power time-increments are anticipated

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

    International Nuclear Information System (INIS)

    Harty, R.B.

    1983-01-01

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

  8. Use of an ions thruster to dispose of type II long-lived fission products into outer space

    International Nuclear Information System (INIS)

    Takahashi, H.; Yu, A.

    1997-01-01

    To dispose of long-lived fission products (LLFPs) into outer space, an ions thruster can be used instead of a static accelerator. The specifications of the ions thrusters which are presently studies for space propulsion are presented, and their usability discussed. Using of a rocket with an ions thruster for disposing of the LLFPs directly into the sun required a larger amount of energy than does the use of an accelerator

  9. 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)

  10. Methods and devices prepared to eliminate activation and fission products from PEC reactor cover gas

    International Nuclear Information System (INIS)

    Caponetti, R.; Gherardi, G.; Petrazzuolo, F.

    1987-01-01

    The major effort made in Italy for the development of fast nuclear reactor is concentrated in the PEC reactor, whose construction is now in the completion stage. The PEC reactor (Prova Elementi di Combustibile - Fuel Element Testing ) is a sodium-cooled reactor with a power rating of 120 MWt, being built for the purpose of studying the behavior of fuel elements under thermal and neutronic conditions similar to those of fast reactor power stations, whit particular attention to safety aspects. The PEC reactor represents a research instrument particularly suitable for studies and experiments in the following fields: performances of the fuel element and its economical optimization (also with the possibility of testing fuel elements not necessarily based on mixed oxides); experiments in the safety field, not only referred to fuel elements, but also to plant subsystems. The experimental program will cover the research of the limit conditions of the typical parameters, such as cladding temperature, linear power, radiation rate, etc. PEC will also allow researches on new-concept fuel elements and thermal, hydraulic and power transients and cycles foreseen in the commercial power plants under normal, upset and emergency conditions. A number of the solutions regarding the PEC reactor and preparatory approaches to its operation are reported in this paper. In particular the following items are discussed: a description of three cover-gas circuits present in the reactor; an estimate of the contamination conditions foreseen under operating conditions; a description of the equipment for the purification of the cover gas and relative operating conditions. There are three cover-gas circuits present in the PEC reactor. They concern the following sodium circuits: primary reactor, primary emergency reactor and sodium purification primary reactor; secondary reactor, test channel and emergency reactor; primary test channel

  11. A Two-Phase Cooling Loop for Fission Surface Power Waste Heat Transport, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Current lunar-based Fission Surface Power (FSP) Systems that will support sustained surface outposts consist of a nuclear reactor with power converters, whose waste...

  12. Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

    Science.gov (United States)

    Clement, J. D.

    1973-01-01

    Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

  13. 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.

  14. 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

  15. Potential for the use of hydrochloric acid in fission reactor fuel recycle

    International Nuclear Information System (INIS)

    Mailen, J.C.; Bell, J.T.

    1985-01-01

    The chemistry and the effects of the use of hydrochloric acid as the aqueous phase in fuel recycle are surveyed. Available data are sufficient to suggest that separations of actinides and fission products can be at least as good in an HCl-trialkyl amine system as in Purex. Advantages of the HCl system are simpler operations of the off-gas system, better separation of neptunium from uranium and plutonium, better control of oxidation states of the dissolved species, and simpler recycle of the acid. A possible advantage is the more complete dissolution of the fission products, leaving very little insoluble residue. Disadvantages include lack of development of methods for dissolution of oxide fuel in hydrochloric acid, the requirement for processing equipment constructed of titanium, possible complications in the waste-handling system, and the dissolution of much of the cladding in the case of stainless-steel clad fuel

  16. Studies on short-lived fission products at the Mainz TRIGA reactor

    International Nuclear Information System (INIS)

    Trautmann, N.

    1974-01-01

    Neutron-rich nuclei of medium mass number are produced by thermal-neutron-induced fission of heavy elements, e.g., 235 U, 239 Pu, and 249 Cf. Pulse irradiations lead to an enhancement of the ratio of short-lived activities to the accompanying longer-lived components. One approach for investigating the properties of short-lived nuclei consists in a combination of rapid chemical separations with higher-resolution gamma spectroscopy. This is demonstrated by the isolation of neutron-rich isotopes of niobium by sorption on glass and of ruthenium by solvent extraction. Other rapid separation procedures from aqueous solutions are briefly summarized and a few examples for their application in nuclear fission- and delayed neutron studies are given. Some experiments with an on-line mass separator of the ISOLDE-type, using chemical targets, are described. (U.S.)

  17. General scheme of research reactor mainly for production of fission 99Mo

    International Nuclear Information System (INIS)

    Shen Feng; Liu Xingmin; Wu Xiaochun; Sun Zheng; Guo Chunqiu; Yi Dayong

    2012-01-01

    On the basis of the analysis for current circumstance and development tendency of research reactor mainly for 99 Mo production in the world, the design idea of this sort of research reactor was proposed. By the optimization and basic design, the general design parameters of the reactor were analyzed and testified. The evaluation of output activities of 99 Mo and the analysis of economics were conducted on the basically assumption. It is argued that the economics of this reactor is improved dramatically while the safety is ensured by the analysis. (authors)

  18. 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

  19. Measurement and analysis of 14 MeV neutron-induced double-differential neutron emission cross sections needed for fission and fusion reactor technology

    International Nuclear Information System (INIS)

    Wang Dahai.

    1990-10-01

    The main objectives of this IAEA Co-ordinated Research Programme are to improve the data on 14 MeV neutron-induced double-differential neutron emission cross sections for materials needed for fission and fusion reactor technology. This report summarizes the conclusions and recommendations which were agreed by all participants during the Second Research Co-ordination Meeting

  20. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

  1. Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pomorski, Michal; Mer-Calfati, Christine [CEA-LIST, Diamond Sensors Laboratory, 91191, Gif-sur-Yvette (France); Foulon, Francois [CEA, National Institute for Nuclear Science and Technology, 91191, Gif-sur-Yvette (France); Sklenka, Lubomir; Rataj, Jan; Bily, Tomas [Department of Nuclear Reactors,Faculty of Nuclear Science and Physical Engineering, Czech Technical University, V. Holesovickach 2, 180 00 PRAHA 8 (Czech Republic)

    2015-07-01

    Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in

  2. Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

    International Nuclear Information System (INIS)

    Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois; Sklenka, Lubomir; Rataj, Jan; Bily, Tomas

    2015-01-01

    Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm 2 . Detectors with surfaces up to 1 cm 2 can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm 2 , with the possibility to enlarge the surface of the detector up to 1 cm 2 . These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The

  3. On-Line Fission Gas Release Monitoring System in the High Flux Reactor Petten

    International Nuclear Information System (INIS)

    Laurie, M.; Fuetterer, M. A.; Appelman, K.H.; Lapetite, J.-M.; Marmier, A.; Knol, S.; Best, J.

    2013-06-01

    For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed dubbed the 'Sweep Loop Facility' (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experimental rigs. The SLF enables continuous and independent surveillance of all gas circuits. The release of volatile fission products (FP) from the in-pile experiments is monitored by continuous gas purging. The fractional release of these FP, defined as the ratio between release rate of a gaseous fission isotope (measured) to its instantaneous birth rate (calculated), is a licensing-relevant test for HTR fuel. The developed gamma spectrometry station allows for higher measurement frequencies, thus enabling follow-up of rapid and massive release transients. The designed stand-alone system was tested and fully used through the final irradiation period of the HFR-EU1 experiment which was terminated on 18 February 2010. Its robustness allowed the set up to be used as extra safety instrumentation. This paper describes the gas activity measurement technique based on HPGe gamma spectrometry and illustrates how qualitative and quantitative analysis of volatile FP can be performed on-line. (authors)

  4. Theoretical Description of the Fission Process

    International Nuclear Information System (INIS)

    Nazarewicz, Witold

    2009-01-01

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation's nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

  5. Study of the HTGR fission product migration at the Osiris experimental reactor

    International Nuclear Information System (INIS)

    Homme, A. l'; Lucot, M.

    1977-01-01

    A program of study on accidents in HTR reactor operation is presented: blowdown of primary coolant circuit, water inlet into the primary circuit, fuel element overheating by pipe logging or loss of cooling. These studies will be made in Aida irradiation loop in the pool of the Osiris reactor [fr

  6. Mirror fusion--fission hybrids

    International Nuclear Information System (INIS)

    Lee, J.D.

    1978-01-01

    The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

  7. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  8. 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

  9. 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)

  10. 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

  11. 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

  12. Irradiation effects on fuels for space reactors

    International Nuclear Information System (INIS)

    Ranken, W.A.; Cronenberg, A.W.

    1984-01-01

    A review of irradiation-induced swelling and gas release experience is presented here for the three principal fuels UO 2 , UC, and UN. The primary advantage of UC and UN over UO 2 is higher thermal conductivity and attendant lower fuel temperature at equivalent pellet diameter and power density, while UO 2 offers the distinct benefit of well-known irradiation performance. Irradiation test results indicate that at equivalent burnup, temperature, and porosity conditions, UC experiences higher swelling than UO 2 or UN. Fission gas swelling becomes important at fuel temperatures above 1320 K for UC, and at somewhat higher temperatures for UO 2 and UN. Evidence exists that at equivalent fuel temperatures and burnups, high density UO 2 and UN experience comparable swelling behavior; however, differences in thermal conductivity influence overall irradiation performance. The low conductivity of UO 2 results in higher thermal gradients which contribute to fuel microcracking and gas release. As a result UO 2 exhibits higher fractional gas release than UN, at least or burnups up to about 3%

  13. 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

  14. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  15. Gamma-ray spectrometric measurements of fission rate ratios between fresh and burnt fuel following irradiation in a zero-power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kröhnert, H., E-mail: hanna.kroehnert@ensi.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Perret, G.; Murphy, M.F. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Chawla, R. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2013-01-11

    The gamma-ray activity from short-lived fission products has been measured in fresh and burnt UO{sub 2} fuel samples after irradiation in a zero-power reactor. For the first time, short-lived gamma-ray activity from fresh and burnt fuel has been compared and fresh-to-burnt fuel fission rate ratios have been derived. For the measurements, well characterized fresh and burnt fuel samples, with burn-ups up to 46 GWd/t, were irradiated in the zero-power research reactor PROTEUS. Fission rate ratios were derived based on the counting of high-energy gamma-rays above 2200 keV, in order to discriminate against the high intrinsic activity of the burnt fuel. This paper presents the measured fresh-to-burnt fuel fission rate ratios based on the {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV) high-energy gamma-ray lines. Comparisons are made with the results of Monte Carlo modeling of the experimental configuration, carried out using the MCNPX code. The measured fission rate ratios have 1σ uncertainties of 1.7–3.4%. The comparisons with calculated predictions show an agreement within 1–3σ, although there appears to be a slight bias (∼3%).

  16. Energy production using fission fragment rockets

    International Nuclear Information System (INIS)

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs

  17. 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.

  18. An Investigation of Power Stabilization and Space-Dependent Dynamics of a Nuclear Fluidized-Bed Reactor

    International Nuclear Information System (INIS)

    Pain, Christopher C.; Eaton, Matthew D.; Gomes, Jefferson L.M.A.; Oliveira, Cassiano R.E. de; Umpleby, Adrian P.; Ziver, Kemal; Ackroyd, Ron T.; Miles, Bryan; Goddard, Antony J.H.; Dam, H. van; Hagen, T.H.J.J. van der; Lathouwers, D.

    2003-01-01

    Previous work into the space-dependent kinetics of the conceptual nuclear fluidized bed has highlighted the sensitivity of fission power to particle movements within the bed. The work presented in this paper investigates a method of stabilizing the fission power by making it less sensitive to fuel particle movement. Steady-state neutronic calculations are performed to obtain a suitable design that is stable to radial and axial fuel particle movements in the bed. Detailed spatial/temporal simulations performed using the finite element transient criticality (FETCH) code investigate the dynamics of the new reactor design. A dual requirement of the design is that it has a moderate power output of ∼300 MW(thermal)

  19. Hardware based technology assessment in support of near-term space fission missions

    International Nuclear Information System (INIS)

    Houts, Mike; Van Dyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky; Salvail, Pat; Williams, Eric; Harper, Roger; Hrbud, Ivana; Carter, Robert

    2003-01-01

    Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and utilized. Successful utilization will most likely occur if frequent, significant hardware-based milestones can be achieved throughout the program. Achieving these milestones will depend on the capability to perform highly realistic non-nuclear testing of nuclear systems. This paper discusses ongoing and potential research that could help achieve these milestones

  20. Hardware Based Technology Assessment in Support of Near-Term Space Fission Missions

    Science.gov (United States)

    Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Martin, James; BraggSitton, Shannon; Carter, Robert; Dickens, Ricky; Salvail, Pat; Williams, Eric; Harper, Roger

    2003-01-01

    Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and utilized. Successful utilization will most likely occur if frequent, significant hardware-based milestones can be achieved throughout the program. Achieving these milestones will depend on the capability to perform highly realistic non-nuclear testing of nuclear systems. This paper discusses ongoing and potential research that could help achieve these milestones.

  1. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. 12-month progress report, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Schultz, K.R.; Baxi, C.B.; Rao, R.

    1976-01-01

    This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

  2. Nuclear Fission Reactor Safety Research in FP7 and future perspectives

    CERN Document Server

    Garbil, Roger

    2014-01-01

    The European Union (ЕU) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

  3. 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

  4. 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

  5. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

  6. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.

    1994-01-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  7. Natural fission reactors in the Franceville basin, Gabon: A review of the conditions and results of a open-quotes critical eventclose quotes in a geologic system

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Holliger, P.; Blanc, P.L.

    1996-01-01

    Natural nuclear fission reactors are only known in two uranium deposits in the world, the Oklo and Bangombe deposits of the Franceville basin: Gabon. Since 1982, five new reactor zones have been discovered in these deposits and studied since 1989 in a cooperative European program. New geological, mineralogical, and geochemical studies have been carried out in order to understand the behavior of the actinides and fission products which have been stored in a geological environment for more than 2.0 Ga years. The Franceville basin and the uranium deposits remained geologically stable over a long period of time. Therefore, the sites of Oklo and Bangombe are well preserved. For the reactors, two main periods of actinide and radionuclides migration have been observed: during the criticality, under P-T conditions of 300 bars and 400-500 degrees C, respectively, and during a distention event which affected the Franceville basin 800 to 900 Ma ago and which was responsible for the intrusion of dolerite dikes close to the reactors. New isotopic analyses on uranium dioxides, clays, and phosphates allow us to determine their respective importance for the retention of fission products. The UO 2 matrix appears to be efficient at retaining most actinides and fission products such as REEs, Y, and Zr but not the volatile fission products (Cd, Cs, Xe, and Kr) nor Rb, Sr, and Ba. Some fissiogenic elements such as Mo, Tc, Ru, Rh, Pd, and Te could have formed metallic and oxide inclusion in the UO 2 matrix which are similar to those observed in artificial spent fuel. Clays and phosphate minerals also appear to have played a role in the retention of fissiogenic REEs and also of Pu. 82 refs., 21 figs., 12 tabs

  8. 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)

  9. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    International Nuclear Information System (INIS)

    Sforza, P.M.; Cresci, R.J.

    1997-01-01

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical

  10. Assessments of sheath strain and fission gas release data from 20 years of power reactor fuel irradiations

    International Nuclear Information System (INIS)

    Purdy, P.L.; Manzer, A.M.; Hu, R.H.; Gibb, R.A.; Kohn, E.

    1997-01-01

    Over the past 20 years, many fuel elements or bundles discharged from Canadian CANDU power reactors have been examined in the AECL hot cells. The post-irradiation examination (PIE) database covers a wide range of operating conditions, from which fuel performance characteristics can be assessed. In the present analysis, a PIE database was compiled representing elements from a total of 129 fuel bundles, of which 26% (34 bundles) were confirmed to have one or more defective elements. This comprehensive database was assessed in terms of measured sheath strain and fission gas release (FGR) for intact elements, in an attempt to identify any changes in these parameters over the history of CANDU reactor operation. Results from this assessment indicate that, for the data that are typical of normal CANDU operating conditions, tensile sheath strain and FGR have remained within 0.5% and 8%, respectively. Those data beyond these ranges are from fuel operated under abnormal conditions, not representative of normal operation, and thus do not indicate a trend toward unexpected fuel behaviour. The distributions of the PIE measurements indicate that maximum expected sheath strains and FGR for normally operated fuel are 0.7% and 13%, respectively. (author)

  11. Average cross-sections for /n, p/ reactions on calcium in a fission-type reactor spectrum

    International Nuclear Information System (INIS)

    Bruggeman, A.; Maenhaut, W.; Hoste, J.

    1974-01-01

    The average cross-section in a fission-type reactor spectrum sigmasub(F) was experimentally determined for the reactions 42 Ca/n,p/ 42 K, 43 Ca/n,p/ 43 K and 44 Ca/n,p/ 44 K. Calcium carbonate samples and fast neutron flux monitors were irradiated with and without cadmium shielding in the Thetis reactor (Institute for Nuclear Sciences, Rijksuniversiteit Gent). The potassium activities induced in the calcium carbonate samples were separated and purified by tetraphenylborate precipitation, after which they were measured with a Ge/Li/-detector of calibrated detection efficiency. On the basis of sigmasub(F)=0.64 mb for the reaction 27 Al/n,α/ 24 Na, the average cross-sections were as follows: 42 Ca/n,p/ 42 K: 2.82+-0.07 mb; 43 Ca/n,p/ 43 K: 1.89+-0.05 mb; 44 Ca/n,p/ 44 K: 0.065+-0.003 mb. (T.G.)

  12. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2002-01-01

    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  13. Fission Surface Power Technology Development Update

    Science.gov (United States)

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

    2011-01-01

    Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power

  14. Nuclear fission sustainability with subcritical reactors driven by external neutron sources

    International Nuclear Information System (INIS)

    Lafuente, A.; Piera, M.

    2011-01-01

    Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.

  15. Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials

    International Nuclear Information System (INIS)

    Phillpot, Simon R.

    2012-01-01

    The work funded from this project has been published in six papers, with two more in draft form, with submission planned for the near future. The papers are: (1) Kinetically-Evolving Irradiation-Induced Point-Defect Clusters in UO 2 by Molecular-Dynamics Simulation; (2) Kinetically driven point-defect clustering in irradiated MgO by molecular-dynamics simulation; (3) Grain-Boundary Source/Sink Behavior for Point Defect: An Atomistic Simulation Study; (4) Energetics of intrinsic point defects in uranium dioxide from electronic structure calculations; (5) Thermodynamics of fission products in UO 2±x ; and (6) Atomistic study of grain boundary sink strength under prolonged electron irradiation. The other two pieces of work that are currently being written-up for publication are: (1) Effect of Pores and He Bubbles on the Thermal Transport Properties of UO2 by Molecular Dynamics Simulation; and (2) Segregation of Ruthenium to Edge Dislocations in Uranium Dioxide.

  16. Macroscopic calculational model of fission gas release from water reactor fuels

    International Nuclear Information System (INIS)

    Uchida, Masaki

    1993-01-01

    Existing models for estimating fission gas release rate usually have fuel temperature as independent variable. Use of fuel temperature, however, often brings an excess ambiguity in the estimation because it is not a rigorously definable quantity as a function of heat generation rate and burnup. To derive a mathematical model that gives gas release rate explicitly as a function of design and operational parameters, the Booth-type diffusional model was modified by changing the character of the diffusion constant from physically meaningful quantity into a mere mathematical parameter, and also changing its temperature dependency into power dependency. The derived formula was found, by proper choice of arbitrary constants, to satisfactorily predict the release rates under a variety of irradiation histories up to a burnup of 60,000 MWd/t. For simple power histories, the equation can be solved analytically by defining several transcendental functions, which enables simple calculation of release rate using graphs. (author)

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

    Science.gov (United States)

    Carré, Frank

    2014-09-01

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

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

    Directory of Open Access Journals (Sweden)

    Carré Frank

    2014-01-01

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

  19. 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)

  20. 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

  1. 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.

  2. Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

    Directory of Open Access Journals (Sweden)

    Jaroszewicz Janusz

    2014-07-01

    Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

  3. 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

  4. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. Addendum 1. Alternate concepts. 12-month progress report addendum, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Schultz, K.R.; Dee, J.B.; Backus, G.A.; Culver, D.W.

    1976-01-01

    During the course of the Mirror Hybrid Fusion-Fission Reactor study several alternate concepts were considered for various reactor components. Several of the alternate concepts do appear to exhibit features with potential advantage for use in the mirror hybrid reactor. These are described and should possibly be investigated further in the future

  5. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2003-01-01

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity

  6. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  7. Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Wielgosz, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with

  8. 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

  9. Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Durand -Smet, R; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    - Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

  10. Technology requirements for fusion--fission reactors based on magnetic-mirror confinement

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    Technology requirements for mirror hybrid reactors are discussed. The required 120-keV neutral beams can use positive ions. The magnetic fields are 8 T or under and can use NbTi superconductors. The value of Q (where Q is the ratio of fusion power to injection power) should be in the range of 1 to 2 for economic reasons relating to the cost of recirculating power. The wall loading of 14-MeV neutrons should be in the range of 1 to 2 MW/m 2 for economic reasons. Five-times higher wall loading will likely be needed if fusion reactors are to be economical. The magnetic mirror experiments 2XIIB, TMX, and MFTF are described

  11. On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Scherer, W.; Brockmann, H.; Drecker, S.; Gerwin, H.; Haas, K.A.; Kugeler, K.; Ohlig, U.; Ruetten, H.J.; Teuchert, E.; Werner, H.; Wolf, L.

    1994-08-01

    Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

  12. Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

    International Nuclear Information System (INIS)

    Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

    2011-01-01

    This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

  13. 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)

  14. Calculation of fission product behavior in a multiple reactor barriers in case of an accident

    International Nuclear Information System (INIS)

    Ezzedin, A. A.; Dakhil, A. S.; Elbaden, S. E.

    2012-12-01

    Radiation protection of the population in case of a reactor accident utilizes reference levels which are based on doses values. Therefore, adequate provisions for effective and timely dose assessment for population in case of accidents at nuclear power plant (NPP) are important. Developing the background for such provisions is the objective of this study. In particular, an exponential model has been developed and utilized to calculate the release rate of the most volatile gaseous materials from different reactor barriers. Calculation has been performed for noble gases (1 33X e, 1 35X e, 1 38X e, 8 5K r, 8 7K r, 8 8K r) and the halogens(1'3 1I , 1 32I , 1 33I , 1'3 4I , 1 35I ). The effective dose rate equivalent is calculations in the nearly stage of a reactor accident. Calculations are performed using the MCNP-4C code. The results are comparable with the final analysis report which utilizes different codes. Results of our calculation shows no excessive dose in populated regions and it is recommended to use secondary containment barrier for highly reduction of the release rate to the environment. (Author)

  15. Study of short-lived fission products with the aid of an isotope separator connected to reactor R2-0

    International Nuclear Information System (INIS)

    Rudstam, G.

    1976-01-01

    This report constitutes a final report on project 74-3289 together with a preliminary report for project 75-3332. These projects have been included in the budget years 1974/75 and 1975/76 as a contribution to the operating costs of reactor R2-0 at Studsvik. The reactor was used for experimental studies on short-lived fission products with OSIRIS isotope-separator equipment. The scientific programme is very broad. It comprises, in the first place, characterisation of fission products (a study of their excitation levels, measurement of decay properties such as half-life and emission of delayed neutrons, determination of neutron energy spectrum, determination of total decay energy, etc.). An important application of this field of research is the determination of decay heat in nuclear fuel. The programme thus comprises research of a fundamental character and applied research. (H.E.G.)

  16. Multiscale Simulation of Thermo-mechancial Processes in Irradiated Fission-reactor Materials.

    Energy Technology Data Exchange (ETDEWEB)

    Simon R. Phillpot

    2012-06-08

    The work funded from this project has been published in six papers, with two more in draft form, with submission planned for the near future. The papers are: (1) Kinetically-Evolving Irradiation-Induced Point-Defect Clusters in UO{sub 2} by Molecular-Dynamics Simulation; (2) Kinetically driven point-defect clustering in irradiated MgO by molecular-dynamics simulation; (3) Grain-Boundary Source/Sink Behavior for Point Defect: An Atomistic Simulation Study; (4) Energetics of intrinsic point defects in uranium dioxide from electronic structure calculations; (5) Thermodynamics of fission products in UO{sub 2{+-}x}; and (6) Atomistic study of grain boundary sink strength under prolonged electron irradiation. The other two pieces of work that are currently being written-up for publication are: (1) Effect of Pores and He Bubbles on the Thermal Transport Properties of UO2 by Molecular Dynamics Simulation; and (2) Segregation of Ruthenium to Edge Dislocations in Uranium Dioxide.

  17. Development of a thermionic-reactor space-power system. Final summary report

    International Nuclear Information System (INIS)

    1973-01-01

    Initial experimental work led to the award of the first AEC thermionic contract on May 1, 1962, for the development of fission heated thermionic cells with an operating life of 10,000 hours or more. Two types of converters were fabricated: (1) electrically heated, and (2) fission heated where the fuel was either uranium carbide or uranium oxide. Competition between GGA and GE was climaxed on July 1, 1970 by the award to GGA of a contract to develop an in-core thermionic reactor. This report is divided into the following: thermionic research, materials technology, thermionic fuel element development, reactor technology, and systems technology

  18. 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.

  19. 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)

  20. 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

  1. Role and use of nuclear theories and models in practical evaluation of neutron nuclear data needed for fission and fusion reactor design and other nuclear applications

    International Nuclear Information System (INIS)

    Prince, A.

    1975-01-01

    A review of the various nuclear models used in the evaluation of neutron nuclear data for fission and fusion reactors is presented. Computer codes embodying the principles of the relevant nuclear models are compared with each other and with experimental data. The regions of validity and limitations of the conceptual formalisms are also included, along with the effects of the numerical procedures used in the codes themselves. Conclusions and recommendations for future demands are outlined.15 tables, 15 figures, 90 references

  2. The role and use of nuclear theories and models in practical evaluation of neutron nuclear data needed for fission and fusion reactor design and other nuclear applications

    International Nuclear Information System (INIS)

    Prince, A.

    1976-01-01

    A review of the various nuclear models used in the evaluation of neutron nuclear data for fission and fusion reactors is presented. Computer codes embodying the principles of the relevant nuclear models are compared with each other and with experimental data. The regions of validity and limitations of the conceptual formalisms are also included, along with the effects of the numerical procedures used in the codes themselves. Conclusions and recommendations for future demands are outlined. (author)

  3. Status of fission yield measurements

    International Nuclear Information System (INIS)

    Maeck, W.J.

    1979-01-01

    Fission yield measurement and yield compilation activities in the major laboratories of the world are reviewed. In addition to a general review of the effort of each laboratory, a brief summary of yield measurement activities by fissioning nuclide is presented. A new fast reactor fission yield measurement program being conducted in the US is described

  4. Potential of incineration of long-life fission products from fission energy system by D-T and D-D fusion reactors

    International Nuclear Information System (INIS)

    Sekimoto, H.; Takashima, H.

    2001-01-01

    The incineration of LLFPs, all of which can not be incinerated with only the fast reactor without isotope separation is studied by employing the DT and DD fusion reactors. The requirement of production of tritium for the DT reactor is severe and the thickness of the blanket should be decreased considerably to incinerate the considerable amount of LLFPs. On the other hand the DD fusion reactor is free from the neutron economy constraint and can incinerate all LLFPs. The pure DD reactor can also show the excellent performance to reduce the first wall loading less than 1 MW/m 2 even for total LLFP incineration. By raising the wall loading to the design limit, the D-D reactor can incinerate the LLFPs from several fast reactors. When the fusion reactor is utilized as an energy producer, plasma confinement is very difficult problem, especially for the D-D reactor compared to the D-T reactor. However, when it is utilized as an incinerator of LLFP, this problem becomes considerably easier. Therefore, the incineration of LLFP is considered as an attractive subject for the D-D reactor. (author)

  5. Potential of incineration of long-life fission products from fission energy system by D-T and D-D fusion reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Takashima, Hiroaki

    1999-01-01

    The incineration of LLFPs, all of which can not be incinerated with only the fast reactor without isotope separation is studied by employing the DT and DD fusion reactors. The requirement of production of tritium for the DT reactor is severe and the thickness of the blanket should be decreased considerably to incinerate the considerable amount of LLFPs. On the other hand the DD fusion reactor is free from the neutron economy constraint and can incinerate all LLFPs. The pure DD reactor can also show the excellent performance to reduce the first wall loading less than 1 MW/m 2 even for total LLFP incineration. By raising the wall loading to the design limit, the D-D reactor can incinerate the LLFPs from several fast reactors. When the fusion reactor is utilized as an energy producer, plasma confinement is very difficult problem, especially for the D-D reactor compared to the D-T reactor. However, when it is utilized as an incinerator of LLFP, this problem becomes considerably easier. Therefore, the incineration of LLFP is considered as an attractive subject for the D-D reactor. (author)

  6. A cost/benefit analysis of commercial fusion-fission hybrid reactor development

    International Nuclear Information System (INIS)

    Kostoff, R.N.

    1983-01-01

    A simple algorithm was developed that allows rapid computation of the ratio R, of present worth of benefits to present worth of hybrid RandD program costs as a function of potential hybrid unit electricity cost savings, discount rate, electricity demand growth rate, total hybrid RandD program cost, and time to complete a demonstration reactor. In the sensitivity study, these variables were assigned nominal values (unit electricity cost savings of 4 mills/k W-hr, discount rate of 4%/year, growth rate of 2.25%/year, total RandD program cost of $20 billion, and time to complete a demonstration reactor of 30 years), and the variable of interest was varied about its nominal value. Results show that R increases with decreasing discount rate and increasing unit electricity savings and ranges from 4 to 94 as discount rateranges from 5 to 3%/year and unit electricity savings range from 2 to 6 mills/k W-hr. R increases with increasing growth rate and ranges from 3 to 187 as growth rate ranges from 1 to 3.5%/year and unit electricity cost savings range from 2 to 6 mills/k W-hr. R attains a maximum value when plotted against time to complete a demonstration reactor. The location of this maximum value occurs at shorter completion times as discount rate increases, and this optimal completion time ranges from 20 years for a discount rate of 4%/year to 45 years for a discount rate of 3%/year

  7. Home brew technetium : clinical scale desktop plasma fusion neutron source to produce Tc99m as an alternative to industrial scale fission reactor sources

    International Nuclear Information System (INIS)

    Bosi, S.G.; Khachan, J.; Oborn, B.M.

    2011-01-01

    Full text: Tc-99m (decay product of Mo-99) accounts for ∼ 90% of world's production of radiopharmaceuticals. Recent unexpected shutdowns of two fission reactors and routine maintenance closures .e created a global shortage of Tc-99m, hence the large global effort to find alternative sources. This project aims to design and produce a novel prototype Mo-99/Tc-99m source. An operational desktop neutron source is available at the University of Sydney, employing a deuterium fusion-plasma to create 2.45 MeV neutrons. These neutrons will be used to activate Mo-98 thin an activation vessel. In one embodiment, the activation vessel contains an aqueous slurry or gel containing Mo-98 which converts to 0-99 upon activation. The decay product Tc-99m could then be milked, similar to existing Tc-99m generators. Monte Carlo will be :ed to assess yield versus size and geometry for various vessel designs. The neutron source filled with deuterium operating at 250 W, produces 3 x 106 neutrons continuously. The neutron flux can be increased ∼ 100-fold if the fill gas is 50% tritium and by another ∼ 100-1000-fold by increasing the power. This is being designed for local use, perhaps on the scale f one or a few hospitals, so the yield would not need to be industrial ;ale as with fission reactor sources. This device is low cost <$300 K) compared with cyclotrons and fission reactors.

  8. Potential for fissile breeding with the fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.

    1976-01-01

    The general features of the mirror reactor design are discussed. Details of the blanket-coil geometry are shown. The inside face of the blanket segments are divided into individual pressure vessels. These submodules contain fissile breeding material located directly behind the first wall, a fusile breeding material behind the fertile breeder, and then coolant inlet and outlet plena. Two blankets are examined and compared in this study. One contains natural uranium plus 7 wt. percent Mo, the second contains thorium metal. The performance of these blankets is discussed

  9. Development of a calculation method for one dimensional kinetic analysis in fission reactors, with feedback effects

    International Nuclear Information System (INIS)

    Paixao, S.B.

    1985-01-01

    The methodology used in the WIGLE3 computer code is studied. This methodology has been applied for the steady-state and transient solutions of the one-dimensional, two-group, diffusion equations in slab geometry, in axial type probelm analysis. It's also studied, based in a WIGLE3 computer code, reactor representative models, considering non-boiling heat transfer. A steady-state program for control rod bank position search- CITER 1D- has been developed. Some criticality research on the proposed system has been done using different control rod bank initial positions, time steps and convergence parameters. (E.G.) [pt

  10. The effect of load factor on fission product decay heat from discharged reactor fuel

    International Nuclear Information System (INIS)

    Davies, B.S.J.

    1978-07-01

    A sum-of-exponentials expression representing the decay heat power following a burst thermal irradiation of 235 U has been used to investigate the effect of load factor during irradiation on subsequent decay heat production. A sequence of random numbers was used to indicate reactor 'on' and 'off' periods for irradiations which continued for a total of 1500 days at power and were followed by 100 days cooling. It was found that for these conditions decay heat is almost proportional to load factor. Estimates of decay heat uncertainty arising from the random irradiation pattern are also given. (author)

  11. A cost and safety superiority of fusion-fission hybrid reactor in China nuclear energy development

    International Nuclear Information System (INIS)

    Pereslavtszev, P.E.; Luan Guishi; Xia Chengang

    1994-08-01

    Considering economy and safety, an optimization model of nuclear energy developing scenarios of China was set up. An objective function to optimize was determined. Three prospective developing scenarios of China nuclear energy system including hybrid reactor were calculated and discussed. In the system which has no fissile material exchange with other system, a smooth developing model has a smooth distribution of inventory of Pu, thus the potential danger of whole nuclear energy system will be decreased. This scheme will improve investment effectiveness. Result shows that the optimization is necessary and the significant profit in cost and safety can be obtained. (5 tabs., 8 figs., 12 refs.)

  12. Economics analysis of fuel cycle cost of fusion–fission hybrid reactors based on different fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2015-01-15

    Highlights: • Economics analysis of fuel cycle cost of FFHRs is carried out. • The mass flows of different fuel cycle strategies are established based on the equilibrium fuel cycle model. • The levelized fuel cycle costs of different fuel cycle strategies are calculated, and compared with current once-through fuel cycle. - Abstract: The economics analysis of fuel cycle cost of fusion–fission hybrid reactors has been performed to compare four fuel cycle strategies: light water cooled blanket burning natural uranium (Strategy A) or spent nuclear fuel (Strategy B), sodium cooled blanket burning transuranics (Strategy C) or minor actinides (Strategy D). The levelized fuel cycle costs (LFCC) which does not include the capital cost, operation and maintenance cost have been calculated based on the equilibrium mass flows. The current once-through (OT) cycle strategy has also been analyzed to serve as the reference fuel cycle for comparisons. It is found that Strategy A and Strategy B have lower LFCCs than OT cycle; although the LFCC of Strategy C is higher than that of OT cycle when the uranium price is at its nominal value, it would become comparable to that of OT cycle when the uranium price reaches its historical peak value level; Strategy D shows the highest LFCC, because it needs to reprocess huge mass of spent nuclear fuel; LFCC is sensitive to the discharge burnup of the nuclear fuel.

  13. Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

    2009-01-01

    The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

  14. 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

  15. Energetics of semi-catalyzed-deuterium, light-water-moderated, fusion-fission toroidal reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.; Towner, H.H.; Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.

    1978-07-01

    The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U 3 Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10 14 cm -3 s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor

  16. Qualification of SiC materials for fusion and fission reactors

    International Nuclear Information System (INIS)

    Ryazanov, Alexander

    2009-01-01

    Ceramic materials such as silicon carbide (SiC) and SiC/SiC composites are both considered, due to their high-temperature strength, pseudo-ductile fracture behavior and low-induced radioactivity, as candidate materials for fusion reactor (test blanket module for ITER) and high temperature gas-cooled reactors (HTGR). The radiation swelling and creep of SiC are very important physical phenomena that determine the radiation resistance of them in these reactors. Other important problem which exists especially in fusion reactor is an effect of accumulation of high concentrations of helium atoms in SiC (up to 15000-20000 at.ppm) due to (n,α) nuclear reaction on physical mechanical properties. An understanding of the physical mechanism of this phenomenon is very important for the investigations of helium atom effect on radiation swelling in SiC. In this report a compilation of non-irradiated and irradiated properties of SiC are provided and analyzed in terms of their application to fusion and high temperature gas cooled reactors. Special topic of this report is oriented on the micro structural changes in chemically vapor-deposited (CVD) high-purity beta-SiC during neutron and ion irradiations at elevated temperatures. The evolutions of various radiation induced defects including dislocation loops, network dislocations and cavities are presented here as a function of irradiation temperature and fluencies. These observations are discussed in relation with such irradiation phenomena in SiC as low temperature swelling and cavity swelling. One of the main difficulties in the radiation damage studies of SiC materials lies in the absence of theoretical models and interpretation of many physical mechanisms of radiation phenomena including the radiation swelling and creep. The point defects in ceramic materials are characterized by the charge states and they can have an effective charge. The internal effective electrical field is formed due to the accumulation of charged point

  17. Fission energy program of the US Department of Energy, FY 1981

    International Nuclear Information System (INIS)

    1980-03-01

    Information is presented concerning the National Energy Plan and fission energy policy; fission energy program management; converter reactor systems; breeder reactor systems; and special nuclear evaluations and systems

  18. Low-temperature thermionics in space nuclear power systems with the safe-type fast reactor

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Yarygin, V.I.; Lazarenko, G.E.; Zabudko, A.N.; Ovcharenko, M.K.; Pyshko, A.P.; Mironov, V.S.; Kuznetsov, R.V.

    2007-01-01

    The potentialities of the use of the low-temperature thermionic converters (TIC) with the emitter temperature ≤ 1500 K in the space nuclear power system (SNPS) with the SAFE-type (Safe Affordable Fission Engine) fast reactor proposed and developed by common efforts of American experts have been considered. The main directions of the 'SAFE-300-TEG' SNPS (300 kW(thermal)) design update by replacing the thermoelectric converters with the low-temperature high-performance thermionic converters (with the barrier index V B ≤ 1.9 eV and efficiency ≥ 10%) meant for a long-term operation (5 years at least) as the components of the SAFE-300-TIC SNPS for a Lunar base have been discussed. The concept of the SNPS with the SAFE-type fast reactor and low-temperature TICs with specific electric power of about 1.45 W/cm 2 as the components of the SAFE-300-TIC system meeting the Nasa's initial requirements to a Lunar base with the electric power demand of about 30 kW(electrical) for robotic mission has been considered. The results, involving optimization and mass-and-size estimation, show that the SAFE-300-TIC system meets the initial requirements by Nasa to the lunar base power supply. The main directions of the system update aimed at the output electric power increase up to 100 kW(electrical) have also been presented. (authors)

  19. 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

  20. 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

  1. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    1993-02-01

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  2. Non-fertile fuels for burning weapons plutonium in thermal fission reactors

    International Nuclear Information System (INIS)

    Lombardi, C.; Mazzola, A.; Vettraino, F.

    1996-01-01

    In the last few years, the excess plutonium disposition has become ever more a topical and critical issue. As a matter of fact, more than 200 MT of plutonium coming from spent fuel reprocessing have been already stockpiled and over the next decade, under the already ratified agreements, another about 200 MT of weapon-grade plutonium are expected to be available from nuclear weapons dismantlement. On this basis, an ever growing plutonium production is no longer the goal and the already stored quantities should be burnt in power reactors by taking care that no new plutonium is generated under irradiation. This new outlook in considering plutonium has led many designers to reassess the Fast Breeder Reactors (FBR) role and shifting from breeder to burner machines perspective. Several solutions for burning plutonium have been so far proposed and discussed from the safeguards, proliferation resistance, environmental safety, technological background, economy and time schedule standpoint. A proposal for plutonium burning in commercial Pressurized Water Reactors (PWR) by using a non-fertile oxide-type fuel consisting of PuO 2 diluted in an inert matrix is reported hereafter. This solution appears to receive an ever growing interest in the nuclear community. In order not to produce new plutonium during irradiation an innovative U-free fuel is being researched, based on an inert matrix which will consist in a mixed compound of inert oxides, such as ZrO 2 , Al2O 3 , MgO, CeO 2 where the plutonium oxide is dispersed in. The matrix will fulfill the following requirements: good chemical compatibility, acceptable thermal conductivity, good nuclear properties, good stability under irradiation, good dissolution resistance. The plutonium relative content will be comparable to that used in MOX fuel. The fuel is expected to be characterized by a high chemical stability (rock-like fuel), so that after discharge from reactor and adequate cooling time, it can be considered a High Level

  3. 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.

  4. 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

  5. Neutron Diffusion in a Space Lattice of Fissionable and Absorbing Materials

    Science.gov (United States)

    Feynman, R. P.; Welton, T. A.

    1946-08-27

    Methods are developed for estimating the effect on a critical assembly of fabricating it as a lattice rather than in the more simply interpreted homogeneous manner. An idealized case is discussed supposing an infinite medium in which fission, elastic scattering and absorption can occur, neutrons of only one velocity present, and the neutron m.f.p. independent of position and equal to unity with the unit of length used.

  6. Analytical treatment of space-energy neutron distribution in reactor cell

    International Nuclear Information System (INIS)

    Stefanovic, D.

    1971-01-01

    Application of the analytical procedure described in this paper for study of fission spectrum evolution in the reactor cell gave results which enable complete understanding of neutron slowing down and transport p recesses and can be used for testing practical approximative methods. Heterogeneous system made of natural uranium and heavy water was used as an example

  7. Retention of activation and fission radionuclides by mallards from the Test Reactor Area radioactive leaching pond

    International Nuclear Information System (INIS)

    Halford, D.K.; Millard, J.B.; Schreckhise, R.G.

    1978-01-01

    Twenty semi-wild mallard ducks were banded, fitted with dorsal and ventral thermoluminescent dosimeter packets, and released on the Test Reactor Area radioactive leaching ponds. Ducks were live captured after 75 days and 145 days on the pond, placed in metabolic cages and whole-body counted periodically for 52 days. Ducks from each group were sacrificed immediately after capture, dissected, and muscle, feather, gut, and liver samples submitted for analyses. The remaining ducks were also sacrificed and dissected after the 52 day counting period. Concentrations of the 17 gamma emitting radionuclides detected at capture and after 52 days of physical and biological decay were compared. Highest mean radionuclide concentrations were found in feathers followed by gut, liver, and muscle. Effective and biological halflives of Zn-65, Cr-51, Cs-134, Cs-137, and Se-75 were determined and compared with data from previous studies. Samples are currently being analyzed for Pu-238, Pu-239-240, Am-241, Cm-242, Cm-244 and Sr-90. Further data analyses will be completed after data collection has terminated

  8. Investigation of (n,γ) reactions in fissionable fluids in a hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

    2014-06-15

    In this study, the effect of (n,γ) reactions in spent fuel-grade (SFG)-PuO{sub 2}, UO{sub 2}, NpO{sub 2} and UCO contents was investigated in a designed hybrid reactor system. In this system, the molten salt-heavy metal mixtures 93 - 85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2 - 10 % UO{sub 2}, 93 - 85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2 - 10 % NpO{sub 2}, 93 - 85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2 - 10 % UCO were used as fluids. Beryllium (Be) is used as neutron multiplier by (n,2n) reactions. Thence, a Be zone with a width of 3 cm was used in order to contribute on fissile fuel breeding between the liquid first wall and blanket. (n,γ) reactions were calculated in the liquid first wall, blanket and shield zones, which SFG-PuO{sub 2}, UO{sub 2}, NpO{sub 2} and UCO contents. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code and nuclear data library ENDF/B-VII.0. (orig.)

  9. Technologies for tritium control in fission reactors moderated with heavy water

    International Nuclear Information System (INIS)

    Ramilo, L.B.; Gomez de Soler, S.M.

    1996-01-01

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  10. Transportation requirements for the disposition of excess weapon plutonium by burning in fission reactors

    International Nuclear Information System (INIS)

    Hovingh, J.; Walter, C.E.

    1996-01-01

    Both the US and Russia are planning to dispose of about 50 Mg of excess weapon plutonium over a 25-year period. One option is to transfer the plutonium to Advanced Light Water (power) Reactors (ALWRs) for use as fuel. Subsequent disposal would then be considered commercial spent fuel. This disposition option, like others, involves the transportation of plutonium in various material forms as it proceeds through various points in the recovery operation. This paper examines both the disposition option and the issues surrounding the transportation of 50 Mg of excess plutonium within the US under current regulatory and infrastructure constraints. Transportation issues include criticality control, shielding, and containment of the contents. Allowable limits on each of these issues are specified by the applicable (or selected) regulation. The composition and form of the radioactive materials to be transported will determine, in part, the applicable portions of the regulations as well as the packaging design. The regulations and the packaging design, along with safeguard and security issues, will determine the quantity of plutonium or fuel assemblies per package as well as the number of packages per shipment and the type of highway carrier. For the disposition of 50 Mg of weapon plutonium using ALWRs in a 25-year campaign, the annual shipment rates are determined for the various types of carriers

  11. Oklo natural fission reactor program. Progress report, April 1-August 31, 1980

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1980-12-01

    An interim report has been published on the redistribution of uranium, thorium, and lead in samples representing several million cubic meters of sandstone and metamorphosed sediments in the Athabasca Basin which is located in the northwest corner of the Canadian province of Saskatchewan. The region of study includes zones of uranium mineralization at Key Lake. Mineralization occurs at the unconformity between the Athabasca sandstone and the underlying metasediments and in fault zones within the metasediments. Lead isotopes record a radiometric age of 1300 +- 150 m.y. in samples from above and below the unconformity. This age probably reflects the time of deposition of the sandstones and an associated redistribution of uranium and/or lead in the underlying rocks. Many of the samples have been fractionated with respect to radiogenic lead and the actinide parent elements since that time. Sandstones and altered rocks from the region above the unconformity have been a transport path and are a repository for lead. In contrast, mineralized rocks are deficient in radiogenic lead and must be an important source of lead in the local geologic environment. Samples from Oklo reactor zone 9 and nearby host rocks have been prepared for isotopic analyses of ruthenium, molybdenum, uranium and lead

  12. Investigation of (n,γ) reactions in fissionable fluids in a hybrid reactor system

    International Nuclear Information System (INIS)

    Guenay, Mehtap

    2014-01-01

    In this study, the effect of (n,γ) reactions in spent fuel-grade (SFG)-PuO 2 , UO 2 , NpO 2 and UCO contents was investigated in a designed hybrid reactor system. In this system, the molten salt-heavy metal mixtures 93 - 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 - 10 % UO 2 , 93 - 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 - 10 % NpO 2 , 93 - 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 - 10 % UCO were used as fluids. Beryllium (Be) is used as neutron multiplier by (n,2n) reactions. Thence, a Be zone with a width of 3 cm was used in order to contribute on fissile fuel breeding between the liquid first wall and blanket. (n,γ) reactions were calculated in the liquid first wall, blanket and shield zones, which SFG-PuO 2 , UO 2 , NpO 2 and UCO contents. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code and nuclear data library ENDF/B-VII.0. (orig.)

  13. 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)

  14. Diversification of 99Mo/99mTc separation: non–fission reactor production of 99Mo as a strategy for enhancing 99mTc availability.

    Science.gov (United States)

    Pillai, Maroor R A; Dash, Ashutosh; Knapp, Furn F Russ

    2015-01-01

    This paper discusses the benefits of obtaining (99m)Tc from non-fission reactor-produced low-specific-activity (99)Mo. This scenario is based on establishing a diversified chain of facilities for the distribution of (99m)Tc separated from reactor-produced (99)Mo by (n,γ) activation of natural or enriched Mo. Such facilities have expected lower investments than required for the proposed chain of cyclotrons for the production of (99m)Tc. Facilities can receive and process reactor-irradiated Mo targets then used for extraction of (99m)Tc over a period of 2 wk, with 3 extractions on the same day. Estimates suggest that a center receiving 1.85 TBq (50 Ci) of (99)Mo once every 4 d can provide 1.48-3.33 TBq (40-90 Ci) of (99m)Tc daily. This model can use research reactors operating in the United States to supply current (99)Mo needs by applying natural (nat)Mo targets. (99)Mo production capacity can be enhanced by using (98)Mo-enriched targets. The proposed model reduces the loss of (99)Mo by decay and avoids proliferation as well as waste management issues associated with fission-produced (99)Mo.

  15. Background and derivation of ANS-5.4 standard fission product release model. Technical report

    International Nuclear Information System (INIS)

    1982-01-01

    ANS Working Group 5.4 was established in 1974 to examine fission product releases from UO2 fuel. The scope of ANS-5.4 was narrowly defined to include the following: (1) Review available experimental data on release of volatile fission products from UO2 and mixed-oxide fuel; (2) Survey existing analytical models currently being applied to lightwater reactors; and (3) Develop a standard analytical model for volatile fission product release to the fuel rod void space. Place emphasis on obtaining a model for radioactive fission product releases to be used in assessing radiological consequences of postulated accidents

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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