Evaluation of aluminum capsules according to ISO 9978 to irradiation of gaseous samples in nuclear reactor

International Nuclear Information System (INIS)

Costa, Osvaldo L. da.; Tiezzi, Rodrigo; Souza, Daiane C.B.; Feher, Anselmo; Moura, Joao A.; Souza, Carla D.; Moura, Eduardo S.; Oliveira, Henrique B.; Zeituni, Carlos A.; Rostelato, Maria Elisa C.M.

2015-01-01

Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of gaseous samples. This paper describes a method to fabricate and evaluate aluminum capsules to irradiate gaseous samples in nuclear reactor. A semi-circular slotted die from a hydraulic press head was modified to seal aluminum tubes. The aluminum capsules were subjected to leak detection tests, which demonstrated the accordance with standard ISO 9978. (author)

Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

International Nuclear Information System (INIS)

Guo, X.; Wehrmeyer, J.A.

1997-01-01

A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-var-epsilon model, RNG k-var-epsilon model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained. copyright 1997 American Institute of Physics

Determination of Kr-85 in environmental samples and gaseous effluents from nuclear industries using the standard method

International Nuclear Information System (INIS)

Heras Iniquez, M.C.; Perez Garcia, M.M.

1983-01-01

The determination of the Kr-85 activity in environmental samples and gaseous effluents from Spanish Nuclear Power Stations is described. The method employed has been published elsewhere. The determinations has been carried out in environmental samples token at JEN Laboratories (Madrid) and the Nuclear Power Stations, Jose Cabrera (Zorita), Garona and Vandellos. Also samples of gaseous effluents of the three plants has been analyzed. Values of the Kr-85 environmental background activity in the Almaraz Nuclear Power Stations, has been determined, before the beginning of its nuclear activity. In this paper the sampling equipment used is described and the values found of Kr-85 activity in all the samples in given. (Author) 29 refs

Research of heat releasing element of an active zone of gaseous nuclear reactor with pumped through nuclear fuel - uranium hexafluoride (UF6)

International Nuclear Information System (INIS)

Batyrbekov, G.; Batyrbekov, E.; Belyakova, E.; Kunakov, S.; Koltyshev, S.

1996-01-01

The purpose of the offered project is learning physics and substantiation of possibility of creation gaseous nuclear reactor with pumped through nuclear fuel-hexafluoride of uranium (Uf6).Main problems of this work are'. Determination of physic-chemical, spectral and optical properties of non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. Research of gas dynamics of laminar, non-mixing two-layer current of gases of hexafluoride of uranium and helium at availability and absence of internal energy release in hexafluoride of uranium with the purpose to determinate a possibility of isolation of hexafluoride of uranium from walls by inert helium. Creation and research of gaseous heat releasing element with pumped through fuel Uf6 in an active zone of research nuclear WWR-K reactor. Objects of a research: Non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. With use of specially created ampoules will come true in-reactor probe and spectral diagnostics of plasma. Calculations of kinetics with the account of main elementary processes proceeding in it, will be carried out. Two-layer non-mixed streams of hexafluoride of uranium and helium at availability and absence of internal energy release. Conditions of obtaining and characteristics of such streams will be investigated. Gaseous heat releasing element with pumped through fuel - Uf6 in an active zone of nuclear WWR-K reactor

Treatment of Radioactive Gaseous Waste

International Nuclear Information System (INIS)

2014-07-01

Radioactive waste, with widely varying characteristics, is generated from the operation and maintenance of nuclear power plants, nuclear fuel cycle facilities, research laboratories and medical facilities. The waste needs to be treated and conditioned as necessary to provide waste forms acceptable for safe storage and disposal. Although radioactive gaseous radioactive waste does not constitute the main waste flow stream at nuclear fuel cycle and radioactive waste processing facilities, it represents a major source for potential direct environmental impact. Effective control and management of gaseous waste in both normal and accidental conditions is therefore one of the main issues of nuclear fuel cycle and waste processing facility design and operation. One of the duties of an operator is to take measures to avoid or to optimize the generation and management of radioactive waste to minimize the overall environmental impact. This includes ensuring that gaseous and liquid radioactive releases to the environment are within authorized limits, and that doses to the public and the effects on the environment are reduced to levels that are as low as reasonably achievable. Responsibilities of the regulatory body include the removal of radioactive materials within authorized practices from any further regulatory control — known as clearance — and the control of discharges — releases of gaseous radioactive material that originate from regulated nuclear facilities during normal operation to the environment within authorized limits. These issues, and others, are addressed in IAEA Safety Standards Series Nos RS-G-1.7, WS-G-2.3 and NS-G-3.2. Special systems should be designed and constructed to ensure proper isolation of areas within nuclear facilities that contain gaseous radioactive substances. Such systems consist of two basic subsystems. The first subsystem is for the supply of clean air to the facility, and the second subsystem is for the collection, cleanup and

Nuclear reactor core catcher

International Nuclear Information System (INIS)

1977-01-01

A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

Nuclear characteristic simulation device for reactor core

International Nuclear Information System (INIS)

Arakawa, Akio; Kobayashi, Yuji.

1994-01-01

In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

Space nuclear power and man's extraterrestrial civilization

International Nuclear Information System (INIS)

Angelo, J.J.; Buden, D.

1983-01-01

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

Control of semivolatile radionuclides in gaseous effluents at nuclear facilities

International Nuclear Information System (INIS)

1982-01-01

An up-to-date review is presented of the subject, combining the results of laboratory studies on control of the most important semivolatile radionuclides in gaseous effluents at nuclear facilities and the results of operating experience in that area. Ruthenium is the most significant semivolatile contaminant in gaseous effluents at nuclear facilities. Volatilization of ruthenium can be reduced by various means, in particular by adding reductants. Volatilized ruthenium can be retained by adsorbents such as silica gel and ferric-oxide-based materials. Decontamination factors in the order of 10 3 have been obtained with these adsorbents under optimum conditions. Volatilized ruthenium can also be removed by other equipment such as condensers and scrubbers. Experience with high-level liquid waste solidification plants has shown that, in general, ruthenium volatilization is in the order of 10% or more unless special treatment is undertaken. There is little experience with ruthenium adsorbers in plants. Silica gel seems to have performed best, with ruthenium decontamination factors of about 10 2 to 10 3 . However, feed-to-stack ruthenium decontamination factors of 10 9 or more have been obtained even without ruthenium adsorbers. Other semivolatiles are relatively insignificant under normal conditions because of a low level of volatilization potential or mass or activity in the inventory. Moreover, owing to particulate formation, they can be easily removed without specific equipment

Nuclear reactor core flow baffling

International Nuclear Information System (INIS)

Berringer, R.T.

1979-01-01

A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

Fluorine nuclear magnetic resonance study of enrichment effects in gaseous, liquid and solid uranium hexafluoride

International Nuclear Information System (INIS)

Ursu, I.; Demco, D.E.; Simplaceanu, V.; Valcu, N.

1977-01-01

The nuclear magnetic resonance method is able to provide information concerning the isotopic content of 235 U in UF 6 by means of measuring the nuclear magnetic transverse relaxation time (T,L2) of 19 F nuclei in liquid UF 6 . In this work, the sources of errors in the T 2 measurements have been analysed and methods for reducing them are dicussed. Typical errors in T 2 determinations are below 2%. The enrichment estimations made by using the linear calibration curves had a deviation of less than 2% with some exceptions. It was found that the chemical impurities may significantly affect the enrichment estimations. 19 F NMR spectra of liquid and gaseous UF 6 at low pressures did not reveal any structure or enrichment effect. The longitudinal nuclear magnetic relaxation of 19 F nuclei in low pressure, gaseous and solid UF 6 showed no enrichment dependence, nor the dipolar relaxation time in solid UF 6 did. (author)

Core catcher for nuclear reactor core meltdown containment

International Nuclear Information System (INIS)

Driscoll, M.J.; Bowman, F.L.

1978-01-01

A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

International Nuclear Information System (INIS)

Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

1991-11-01

Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

Magnetic nuclear core restraint and control

International Nuclear Information System (INIS)

Cooper, M.H.

1979-01-01

A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

Magnetic nuclear core restraint and control

International Nuclear Information System (INIS)

Cooper, M.H.

1979-01-01

A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

Heterogeneous gas core reactor

International Nuclear Information System (INIS)

Diaz, N.J.; Dugan, E.T.

1983-01-01

A heterogeneous gas core nuclear reactor is disclosed comprising a core barrel provided interiorly with an array of moderator-containing tubes and being otherwise filled with a fissile and/or fertile gaseous fuel medium. The fuel medium may be flowed through the chamber and through an external circuit in which heat is extracted. The moderator may be a fluid which is flowed through the tubes and through an external circuit in which heat is extracted. The moderator may be a solid which may be cooled by a fluid flowing within the tubes and through an external heat extraction circuit. The core barrel is surrounded by moderator/coolant material. Fissionable blanket material may be disposed inwardly or outwardly of the core barrel

Magnetic nuclear core restraint and control

International Nuclear Information System (INIS)

Cooper, M.H.

1978-01-01

Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

Design of radiation shields in nuclear reactor core

International Nuclear Information System (INIS)

Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

2008-01-01

This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

Nuclear core catchers

International Nuclear Information System (INIS)

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1976-01-01

A receptacle is described for taking the molten fragments of a nuclear reactor during a reactor core fusion accident. The receptacle is placed under the reactor. It includes at least one receptacle for the reactor core fragments, with a dome shaped part to distribute the molten fragments and at least one outside layer of alumina bricks around the dome. The characteristic of this receptacle is that the outer layer of bricks contains neutron poison rods which pass through the bricks and protrude in relation to them [fr

Nuclear criticality safety evaluation of Spray Booth Operations in X-705, Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

Sheaffer, M.K.; Keeton, S.C.

1993-01-01

This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested

Application of gaseous core reactors for transmutation of nuclear waste

Science.gov (United States)

Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

1976-01-01

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Nuclear reactor core stabilizing arrangement

International Nuclear Information System (INIS)

Jabsen, F.S.

1976-01-01

A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

Nuclear reactor core servicing apparatus

International Nuclear Information System (INIS)

Andrea, C.

1977-01-01

Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

Gaseous waste processing device in nuclear power plant

International Nuclear Information System (INIS)

Takechi, Eisuke; Matsutoshi, Makoto.

1978-01-01

Purpose: To arrange the units of waste processing devices in a number one more than the number thereof required for a plurality of reactors, and to make it usable commonly as a preliminary waste processing device thereby to effectively use all the gaseous waste processing devices. Constitution: A gaseous waste processing device is constituted by an exhaust gas extractor, a first processing device, a second processing device and the like, which are all connected in series. Upon this occasion, devices from the exhaust gas extractor to the first processing device and valves, which are provided in each of reactors, are arranged in series, on one hand, but valves at the downstream side join one another by one pipeline, and are connected to a stack through a total gaseous waste processing device, on another. (Yoshihara, H.)

Nuclear criticality safety evaluation of large cylinder cleaning operations in X-705, Portsmouth Gaseous diffusion Plant

International Nuclear Information System (INIS)

Sheaffer, M.K.; Keeton, S.C.; Lutz, H.F.

1995-06-01

This report evaluates nuclear criticality safety for large cylinder cleaning operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current cleaning procedures and required hardware/equipment is presented, and documentation for large cylinder cleaning operations is identified and described. Control parameters, design features, administrative controls, and safety systems relevant to nuclear criticality are discussed individually, followed by an overall assessment based on the Double Contingency Principle. Recommendations for enhanced safety are suggested, and issues for increased efficiency are presented

Nuclear reactors with auxiliary boiler circuit

International Nuclear Information System (INIS)

George, B.V.; Cook, R.K.

1976-01-01

A gas-cooled nuclear reactor has a main circulatory system for the gaseous coolant incorporating one or more main energy converting units, such as gas turbines, and an auxiliary circulatory system for the gaseous coolant incorporating at least one steam generating boiler arranged to be heated by the coolant after its passage through the reactor core to provide steam for driving an auxiliary steam turbine, such an arrangement providing a simplified start-up procedure also providing emergency duties associated with long term heat removal on reactor shut down

Nuclear core baffling apparatus

International Nuclear Information System (INIS)

Cooper, F.W. Jr.; Silverblatt, B.L.; Knight, C.B.; Berringer, R.T.

1979-01-01

An apparatus for baffling the flow of reactor coolant fluid into and about the core of a nuclear reactor is described. The apparatus includes a plurality of longitudinally aligned baffle plates with mating surfaces that allow longitudinal growth with temperature increases while alleviating both leakage through the aligned plates and stresses on the components supporting the plates

On-line measurement of gaseous iodine species during a PWR severe accident

Energy Technology Data Exchange (ETDEWEB)

Haykal, I.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, 91191 Gif sur Yvette Cedex, (France); Perrin, A. [CNRS-University of Paris Est and Paris 7, Laboratoire Inter-Universitaire des Systemes Atmospheriques, 94010 Creteil, (France); Vincent, B. [University of Burgundy, Laboratoire de physique, CNRS UMR 5027, 9, Avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Manceron, L. [Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); Mejean, G. [University of Joseph Fourier in Grenoble, Laboratoire de Spectrometrie Physique-CNRS UMR 5588, 38402 Saint Martin d' Heres, (France); Ducros, G. [CEA Cadarache, CEA, DEN, Departement d' Etudes des Combustibles, 13108 Saint-Paul-lez-Durance cedex, (France)

2015-07-01

A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

Method for refuelling a nuclear reactor core

International Nuclear Information System (INIS)

Anon.

1977-01-01

This invention relates to an improved method for refuelling a nuclear reactor core inside a reactor vessel. The technique allows a substantial reduction in the refuelling time as compared with previously known methods and permits fewer out of core operations and smaller temporary storage space. (U.K.)

Decree No 74-945 of 6 November 1974 concerning gaseous radioactive effluent releases from large nuclear installations and nuclear installations located on the same site

International Nuclear Information System (INIS)

1974-01-01

This Decree prescribes the licensing for the release of gaseous wastes from nuclear installations as well as the technical supervision of such operations. It does not apply to the transport of radioactive effluents which is governed by the regulations on the transport of dangerous goods. (NEA) [fr

Introduction to the nuclear criticality safety evaluation of facility X-705, Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

Sheaffer, M.K.; Keeton, S.C.

1993-01-01

This report is the first in a series of documents that will evaluate nuclear criticality safety in the Decontamination and Recovery Facility, X-705, Portsmouth Gaseous Diffusion Plant. It provides an overview of the facility, categorizes its functions for future analysis, reviews existing NCS documentation, and explains the follow-on effort planned for X-705. A detailed breakdown of systems, subsystems, and operational areas is presented and cross-referenced to existing NCS documentation

Evaluating nuclear physics inputs in core-collapse supernova models

Science.gov (United States)

Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

Gaseous products and smoke generation on combustion of the insulation materials of nuclear cables

International Nuclear Information System (INIS)

Noguchi, Isamu; Takami, Hiroshi; Ueyama, Michio; Fujimura, Shun-ichi.

1976-01-01

Serious requirements have been introduced to the cables used for nuclear power plants on their flame retardation in the IEEE Standard 383-1974. The movements that the users prescribe the quantity of corrosive gas generated from cables are also observed. This report describes on the measured results of the gaseous products generated by burning polyethylene, polyvinyl-chloride (PVC) and their flame-resistant products, and a part of the covering materials of the cables for nuclear power plants (flame-resistant, crosslinking polyethylene, flame-resistant, low hydrochloric acid PVC, flame-resistant jute) in the infra-red rapid heating combustion test facility designed by the Furukawa Electric Co. Ltd. In addition, the report introduces the test method for the smoke generation evaluation of polymers and a part of the measured results. The gaseous products of combustion were collected and determined quantitatively by gas chromatographic method. Since smoke generation is affected greatly by the kinds, shape, atmosphere, temperature, ignition procedure and others of burnt matters, the establishment of the evaluation test method is difficult, and a number of methods have been proposed. As the measured results showed, it is clear that smoke generation increases with the increase of flame resistant reagent addition. The smoke generation of PVC was of course great in quantity because it contains considerable amount of chlorine for its molecular structure. Flame-resistant polyethylene generates smoke much more than polyethylene without flame-resisting treatment because of its flame resistivity, but less than that of PVC. (Wakatsuki, Y.)

Station blackout core damage frequency in an advanced nuclear reactor

International Nuclear Information System (INIS)

Carvalho, Luiz Sergio de

2004-01-01

Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

Modeling of the core of Atucha II nuclear power plant

International Nuclear Information System (INIS)

Blanco, Anibal

2007-01-01

This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

Improving the calculated core stability by the core nuclear design optimization

International Nuclear Information System (INIS)

Partanen, P.

1995-01-01

Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

Calculation of releases of radioactive materials in gaseous effluents from nuclear-powered merchant ships (NMS-GEFF code)

International Nuclear Information System (INIS)

Cardile, F.P.; Bangart, R.L.; Collins, J.T.

1978-06-01

The Intergovernmental Maritime Consultative Organization IMCO) is currently preparing guidelines concerning the safety of nuclear-powered merchant ships. An important aspect of these guidelines is the determination of the releases of radioactive material in effluents from these ships and the control exercised by the ships over these releases. To provide a method for the determination of these releases, the NRC staff has developed a computerized model, the NMS-GEFF Code, which is described in the following chapters. The NMS-GEFF Code calculates releases of radioactive material in gaseous effluents for nuclear-powered merchant ships using pressurized water reactors

Applications of plasma core reactors to terrestrial energy systems

International Nuclear Information System (INIS)

Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

Preparation of ZnS@In2S3 Core@shell Composite for Enhanced Photocatalytic Degradation of Gaseous o-Dichlorobenzene under Visible Light.

Science.gov (United States)

Liu, Baojun; Hu, Xia; Li, Xinyong; Li, Ying; Chen, Chang; Lam, Kwok-Ho

2017-11-27

In this study, novel ZnS@In 2 S 3 core@shell hollow nanospheres were fabricated by a facile refluxing method for the first time, and the formation mechanism of hollow structure with interior architecture was discussed based on ion-exchange Ostwald ripening. As the photocatalytic material for degradation of gaseous o-Dichlorobenzene (o-DCB), the as-synthesized core@shell hollow nanospheres were found to show significantly enhanced catalytic performance for effective separation of photo-generated charges. Moreover, the mechanisms of enhanced activity were elucidated by band alignment and unique configuration. Such photocatalyst would meet the demands for the control of persistent organic pollutant (POPs) in the atmospheric environment.

Thermal barrier and support for nuclear reactor fuel core

International Nuclear Information System (INIS)

Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

1987-01-01

A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

Real-time simulation of ex-core nuclear instrumentation system

International Nuclear Information System (INIS)

Zhao Qiang; Zhang Zhijian; Cao Xinrong

2005-01-01

Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant (NPP) full-scope training simulator. The simulation method, which is based upon the theory of measurement, is introduced in the paper. The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established. The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration. The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator. (authors)

Fast breeder physics and nuclear core design

International Nuclear Information System (INIS)

Marth, W.; Schroeder, R.

1983-07-01

This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

Apparatus for controlling nuclear core debris

Science.gov (United States)

Jones, Robert D.

1978-01-01

Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

Sensors for use in nuclear reactor cores

International Nuclear Information System (INIS)

Brown, W.L.; Geronime, R.L.

1978-01-01

Sensors including radiation detectors and the like for use within the core of nuclear reactors and which are constructed in a manner to provide optimum reliability of the sensor during use are described

Gaseous diffusion -- the enrichment workhorse

International Nuclear Information System (INIS)

Shoemaker, J.E. Jr.

1984-01-01

Construction of the first large-scale gaseous diffusion facility was started as part of the Manhattan Project in Oak Ridge, Tennessee, in 1943. This facility, code named ''K-25,'' began operation in January 1945 and was fully on stream by September 1945. Four additional process buildings were later added in Oak Ridge as the demand for enriched uranium escalated. New gaseous diffusion plants were constructed at Paducah, Kentucky, and Portsmouth, Ohio, during this period. The three gaseous diffusion plants were the ''workhorses'' which provided the entire enriched uranium demand for the United States during the 1950s and 1960s. As the demand for enriched uranium for military purposes decreased during the early 1960s, power to the diffusion plants was curtailed to reduce production. During the 1960s, as plans for the nuclear power industry were formulated, the role of the diffusion plants gradually changed from providing highly-enriched uranium for the military to providing low-enriched uranium for power reactors

Nuclear clustering - a cluster core model study

International Nuclear Information System (INIS)

Paul Selvi, G.; Nandhini, N.; Balasubramaniam, M.

2015-01-01

Nuclear clustering, similar to other clustering phenomenon in nature is a much warranted study, since it would help us in understanding the nature of binding of the nucleons inside the nucleus, closed shell behaviour when the system is highly deformed, dynamics and structure at extremes. Several models account for the clustering phenomenon of nuclei. We present in this work, a cluster core model study of nuclear clustering in light mass nuclei

Legal provisions governing gaseous effluents radiological monitoring

International Nuclear Information System (INIS)

Winkelmann, I.

1985-01-01

This contribution explains the main provisions governing radiological monitoring of gaseous effluents from LWR type nuclear power plants. KTA rule 1503.1 defines the measuring methods and tasks to be fulfilled by reactor operators in order to safeguard due monitoring and accounting of radioactive substances in the plants' gaseous effluents. The routine measurements are checked by a supervisory programme by an independent expert. The routine controls include analysis of filter samples, comparative measurement of radioactive noble gases, interlaboratory comparisons, and comparative evaluation of measured values. (DG) [de

Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

2007-07-01

Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

Nuclear reactor core modelling in multifunctional simulators

International Nuclear Information System (INIS)

Puska, E.K.

1999-01-01

The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

Nuclear reactor core modelling in multifunctional simulators

Energy Technology Data Exchange (ETDEWEB)

Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

1999-06-01

The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

Spring unit especially intended for a nuclear reactor core

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, Wilhelm.

1977-01-01

This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

The online simulation of core physics in nuclear power plant

International Nuclear Information System (INIS)

Zhao Qiang

2005-01-01

The three-dimensional power distribution in core is one of the most important status variables of nuclear reactor. In order to monitor the 3-D in core power distribution timely and accurately, the online simulation system of core physics was designed in the paper. This system combines core physics simulation with the data, which is from the plant and reactor instrumentation. The design of the system consists of the hardware part and the software part. The online simulation system consists of a main simulation computer and a simulation operation station. The online simulation system software includes of the real-time simulation support software, the system communication software, the simulation program and the simulation interface software. Two-group and three-dimensional neutron kinetics model with six groups delayed neutrons was used in the real-time simulation of nuclear reactor core physics. According to the characteristics of the nuclear reactor, the core was divided into many nodes. Resolving the neutron equation, the method of separate variables was used. The input data from the plant and reactor instrumentation system consist of core thermal power, loop temperatures and pressure, control rod positions, boron concentration, core exit thermocouple data, Excore detector signals, in core flux detectors signals. There are two purposes using the data, one is to ensure that the model is as close as the current actual reactor condition, and the other is to calibrate the calculated power distribution. In this paper, the scheme of the online simulation system was introduced. Under the real-time simulation support system, the simulation program is being compiled. Compared with the actual operational data, the elementary simulation results were reasonable and correct. (author)

Determination of Kr-85 in environmental samples and gaseous effluents from nuclear industries using the standard method; Aplicacion del metodo de concentracion separacion y medida radiactiva por centelleo liquido de Kr-85 en muestras ambientales y en efluentes gaseosos de la industria nuclear

Energy Technology Data Exchange (ETDEWEB)

Heras, M c; Perez, M M

1983-07-01

The determination of the Kr-85 activity in environmental samples and gaseous effluents from Spanish Nuclear Power Stations is described. The method employed has been published elsewhere. The determinations has been carried out in environmental samples token at JEN Laboratories (Madrid) and the Nuclear Power Stations, Jose Cabrera (Zorita), Garona and Vandellos. Also samples of gaseous effluents of the three plants has been analyzed. Values of the Kr-85 environmental background activity in the Almaraz Nuclear Power Stations, has been determined, before the beginning of its nuclear activity. In this paper the sampling equipment used is described and the values found of Kr-85 activity in all the samples in given. (Author) 29 refs.

Real-time advanced nuclear reactor core model

International Nuclear Information System (INIS)

Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

1990-01-01

The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

Protective action evaluation, Part 1. Effectiveness of sheltering as a protective action against nuclear accidents involving gaseous releases

International Nuclear Information System (INIS)

Anno, G.H.; Dore, M.A.

1978-04-01

In an airborne release of radioactive material from a nuclear power plant accident, sheltering of individuals is of importance in emergency protective action planning. An analysis to estimate the effectiveness or benefit that might be derived from sheltering is described. The objective of this effort is the development of sheltering effectiveness information for those responsible for formulating required emergency plans for nuclear power plant siting. Shelter effectiveness is specifically defined as the dose reduction factor (DRF). DRF estimates for different conditions of source release, shelter structure assumptions, and operational time parameters are made for both whole-body and thyroid doses separately, based on a single-compartment structural model of the time-varying outside and inside gaseous radionuclide sources of krypton, zenon, and iodine. Design basis accident (DBA) assumptions are made for the gaseous radionuclide release. The magnitude of the release and dose estimates are based on radionuclide data from The Reactor Safety Study (WASH-1400). Source release time and duration assumptions are related to release categories PWR 1, PWR 3, and PWR 4, for which release times range from 1.5 to 2.5 hr and the release duration ranges from 0.5 to 3 hr. The basic shelter model characteristics considered are gamma ray attentuation, source geometry, gaseous fission-product ingress, and air change rate. Temporal parameters considered are source release time and duration, cloud travel time, and time spent in the shelter structure. Also, the analysis of shelter effectiveness is based on a time-frame model, which can be conveniently related to other operational times important for emergency planning. In addition to developing shelter-effectiveness estimates parametrically, the advantage of exiting and evacuating the vicinity of the shelter area after some initial time in the shelter is analyzed from the standpoint of the DRF and temporal considerations

Order of 4 august 1989 on licensing the release of gaseous radioactive effluents by the Cattenom nuclear production centre (units 3 and 4)

International Nuclear Information System (INIS)

1989-08-01

This Order fixes the conditions and limits of authorised releases of gaseous radioactive effluents from Units 3 and 4 of the Cattenom nuclear power plant. It specifies these are maximum limits, below which the radioactive releases should be as low as possible [fr

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.

In-core fuel management for nuclear reactor

International Nuclear Information System (INIS)

Ross, M.F.; Visner, S.

1986-01-01

This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

Nuclear reactor core safety device

International Nuclear Information System (INIS)

Colgate, S.A.

1977-01-01

The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

Gas core nuclear rocket feasibility project

International Nuclear Information System (INIS)

Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

1997-09-01

The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas core nuclear rocket (GCNR) has the potential to be such a system. The gas core concept relies on the use of fluid dynamic forces to create and maintain a vortex. The vortex is composed of a fissile material which will achieve criticality and produce high power levels. By radiatively coupling to the surrounding fluids, extremely high temperatures in the propellant and, thus, high specific impulses can be generated. The ship velocities enabled by such performance may allow a 9 month round trip, manned Mars mission to be considered. Alternatively, one might consider slightly longer missions in ships that are heavily shielded against the intense Galactic Cosmic Ray flux to further reduce the radiation dose to the crew. The current status of the research program at the Los Alamos National Laboratory into the gas core nuclear rocket feasibility will be discussed

Supporting system for the core restraint of nuclear reactors

International Nuclear Information System (INIS)

Kaser, A.

1973-01-01

The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

Core management and fuel handling for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

Core management and fuel handling for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

IAEA decadal activities in the field of radioactive gaseous waste management

International Nuclear Information System (INIS)

Plumb, G.R.

1991-01-01

The IAEA has long recognized that gaseous waste management is vital in the design and safe operation of all nuclear facilities such that in the decade of the 1980's the IAEA program covered the important aspects of the entire field. The activities reviewed in this paper were marked at the outset by a comprehensive international symposium on the subject in February 1980 organized by the IAEA jointly with the Nuclear Energy Agency of the OECD when the detailed state-of-the-art was established in 43 papers. In the interim, experts have been convened in IAEA sponsored meetings to result in sixteen technical documents which included summaries of three substantial Co-ordinated Research Programs. Early IAEA activities paid particular attention to management of gas radionuclides which from a matured nuclear industry, could be judged to build-up to long-term sources of irradiation for regional and global populations. Mid-term ongoing activities in handling and retention of gaseous radionuclides arising from abnormal operations in nuclear power plants were given much emphasis following the Chernobyl accident. In the latter years the IAEA activities included detailed examinations of the design and operation of gas cleaning systems for the range of nuclear facilities. Technical reports on gaseous waste management were issued relating to high-level liquid waste conditioning plants (including control of semi-volatiles), nuclear power plants, low- and intermediate-level radioactive materials handling facilities and radioactive waste incinerators

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).

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)

Neutronic analysis of the ford nuclear reactor leu core

International Nuclear Information System (INIS)

Raza, S.S.; Hayat, T.

1989-08-01

Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

Preliminary concept of a zero power nuclear reactor core

International Nuclear Information System (INIS)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

2011-01-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Preliminary concept of a zero power nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Nuclear-fuel-cycle education: Module 5. In-core fuel management

International Nuclear Information System (INIS)

Levine, S.H.

1980-07-01

The purpose of this project was to develop a series of educational modules for use in nuclear-fuel-cycle education. These modules are designed for use in a traditional classroom setting by lectures or in a self-paced, personalized system of instruction. This module on in-core fuel management contains information on computational methods and theory; in-core fuel management using the Virginia Polytechnic Institute and State University computer modules; pressurized water reactor in-core fuel management; boiling water reactor in-core fuel management; and in-core fuel management for gas-cooled and fast reactors

Status of core nuclear design technology for future fuel

International Nuclear Information System (INIS)

Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

1997-01-01

The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

2008-07-01

A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

Improvements to the sodium supply system of a nuclear reactor core

International Nuclear Information System (INIS)

Chevallier, Rene; Marchais, Christian.

1981-01-01

This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

Directory of Open Access Journals (Sweden)

Andrea Cerutti

Full Text Available Hepatitis C virus (HCV infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS, but no nuclear export signal (NES has yet been identified.We show here that the aa(109-133 region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126 in the identified NES or in the sequence encoding the mature core aa(1-173 significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

Science.gov (United States)

Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

2011-01-01

Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified.We show here that the aa(109-133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1-173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

Study on in-core fuel management for CNP1500 nuclear power plant

International Nuclear Information System (INIS)

Li Dongsheng

2005-10-01

CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

An introduction to technetium in the gaseous diffusion cascades

International Nuclear Information System (INIS)

Simmons, D.W.

1996-09-01

The radioisotope technetium-99 ( 99 Tc) was introduced into the gaseous diffusion plants (GDP) as a contaminant in uranium that had been reprocessed from spent nuclear reactor fuel. 99 Tc is a product of the nuclear fission of uranium-235 ( 235 U). The significantly higher emitted radioactivity of 99 Tc generates concern in the enrichment complex and warrants increased attention (1) to the control of all site emissions, (2) to worker exposures and contamination control when process equipment requires disassembly and decontamination, and (3) to product purity when the enriched uranium hexafluoride (UF 6 ) product is marketed to the private sector. A total of 101,268 metric tons of RU (∼96% of the total) was fed at the Paducah Gaseous Diffusion Plant (PGDP) between FY1953 and FY1976. An additional 5600 metric tons of RU from the government reactors were fed at the Oak Ridge Gaseous Diffusion Plant (ORGDP), plus an approximate 500 tons of foreign reactor returns. Only a small amount of RU was fed directly at the Portsmouth Gaseous Diffusion Plant (PORTS). The slightly enriched PGDP product was then fed to either the ORGDP or PORTS cascades for final enrichment. Bailey estimated in 1988 that of the 606 kg of Tc received at PGDP from RU, 121 kg was subsequently re-fed to ORGDP and 85 kg re-fed to PORTS

Emergency core cooling systems in CANDU nuclear power plants

International Nuclear Information System (INIS)

1981-12-01

This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

Core support structure for nuclear power plants

International Nuclear Information System (INIS)

Steinkamp, E.; Tautz, J.; Ries, H.

1979-01-01

A core support structure for nuclear power plants includes a grid of mutually crossing bridges and a support ring surrounding the grid and connected to ends of the outer bridges of the grid, the grid being formed of profile rod crosses having legs of given length, respective legs of pairs of adjacent crosses abutting one another endwise to form together a side of the smallest mesh opening of the grid, and weld means for securing the profile rod crosses to one another at the mutually abutting ends of the legs thereof; and method of producing the foregoing core support structure

Nuclear-powered rocket of the future

Energy Technology Data Exchange (ETDEWEB)

Yunqiao, B

1979-06-01

A possible manned mission to Mars with a crew of 7 in an 80-meter-long nuclear-powered rocket will take 180 days to reach its destination, will spend 10 to 14 days on the surface, and will take 200 days to return. A nuclear-powered engine (using U-235 or U-239) is the most likely means of propulsion. Four designs are described. The superheated exhaust engine will use a reactor to heat liquid hydrogen to over 4000/sup 0/C, after which it will be ejected from the exhaust. A plasma compression engine will use electric current produced by a reactor to heat hydrogen to plasma temperature (70,000/sup 0/C), after which it will be ejected through the exhaust by a magnetic field. In a gaseous-core reactor engine, gaseous fuel will heat liquid hydrogen to over 9,000/sup 0/C and use it as the propellant. The boldest solution is a proposal to use small nuclear explosions as the propulsive force. The first alternative will probably not produce enough thrust, while there will be a difficulty producing sufficient electricity in the second alternative. The other two alternatives seem promising.

In-core gamma dosimetry by solid state nuclear track detectors

International Nuclear Information System (INIS)

Khan, H.A.

1980-02-01

Results are reported of a study undertaken to develop Solid State Nuclear Track Detectors (SSNTD) for the measurement of gamma doses in the megarad region such as those existing in and around a nuclear reactor core. The changes brought about in the track etching parameters and in the ultraviolet and infrared transmittances, have been studied for possible use as gamma dose measuring indices. Effects of various parameters in the core such as neutron flux, beta particles, water, temperature, and gamma ray spectrum have been investigated and found to have only small influence on the proposed gamma dose measuring indices

A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

International Nuclear Information System (INIS)

Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

2008-01-01

The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

Energy Technology Data Exchange (ETDEWEB)

Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

2008-09-15

The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

International Nuclear Information System (INIS)

Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

2009-01-01

The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

Nuclear reactor core and fuel element therefor

International Nuclear Information System (INIS)

Fortescue, P.

1986-01-01

This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments

International Nuclear Information System (INIS)

Liu, Hanying; Miller, Don W.; Li, Dongxu; Radcliff, Thomas D.

2002-01-01

For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then it would play a valuable and critical role in increasing nuclear power, predicting abnormal reactor operation, improving core physical models and reducing core thermal margin so as to implement higher fuel burn-up. Furthermore, the management of core thermal margin and fuel operation may be easier during reactor operation, post-accident or spent fuel storage. On the other hand, for some advanced Generation IV reactors, the sealed and long-lived reactor core design challenges traditional measurement techniques while conventional ex-core detectors and current in-core detectors can not monitor details of the in-core fuel conditions. A method is introduced in this paper that responds to the challenge to measure nuclear power and to monitor the in-core thermal environments, for example, local fuel pin or coolant heat convection coefficient and temperature. In summary, the method, which has been designed for online in-core measurement and surveillance, will be beneficial to advanced plant safety, efficiency and economics by decreasing thermal margin or increasing nuclear power. The method was originally developed for a constant temperature power sensor (CTPS). The CTPS is undergoing design and development for an advanced reactor core to measure in-core nuclear power in measurement mode and to monitor thermal environments in compensation mode. The sensor dynamics was analyzed in compensation mode to determine the environmental temperature and the heat transfer coefficient. Previous research demonstrated that a first order dynamic model is not sufficient to simulate sensor

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

Phase diagram of nuclear 'pasta' and its uncertainties in supernova cores

International Nuclear Information System (INIS)

Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Yasuoka, Kenji; Ebisuzaki, Toshikazu

2008-01-01

We examine the model dependence of the phase diagram of inhomogeneous nulcear matter in supernova cores using the quantum molecular dynamics (QMD). Inhomogeneous matter includes crystallized matter with nonspherical nuclei--''pasta'' phases--and the liquid-gas phase-separating nuclear matter. Major differences between the phase diagrams of the QMD models can be explained by the energy of pure neutron matter at low densities and the saturation density of asymmetric nuclear matter. We show the density dependence of the symmetry energy is also useful to understand uncertainties of the phase diagram. We point out that, for typical nuclear models, the mass fraction of the pasta phases in the later stage of the collapsing cores is higher than 10-20%

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.

Modular core component support for nuclear reactor

International Nuclear Information System (INIS)

Finch, L.M.; Anthony, A.J.

1975-01-01

The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

Nuclear detectors for in-core power-reactors

International Nuclear Information System (INIS)

Duchene, Jean; Verdant, Robert.

1979-12-01

Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

Demister apparatus for gaseous wastes carrying radioactive aerosols

International Nuclear Information System (INIS)

Meline, F.G.; Richter, R.J.

1983-01-01

In the nuclear industry, more precisely in the field of spent fuel reprocessing, the cleaning of the gaseous wastes, before evacuation, should be realized with a device designed in order to take full account of the constraints that are inherent in the radioactive media. The French Atomic Authority (CEA), in collaboration with the Societe Generale pour les Techniques Nouvelles, have studied and developed types of demister for the nuclear field having good cleaning properties

A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks

International Nuclear Information System (INIS)

Lima, Alan M.M. de; Schirru, Roberto

2005-01-01

A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

Impact of nuclear 'pasta' on neutrino transport in collapsing stellar cores

International Nuclear Information System (INIS)

Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Takiwaki, Tomoya; Yasuoka, Kenji; Ebisuzaki, Toshikazu

2007-01-01

Nuclear 'pasta', nonspherical nuclei in dense matter, is predicted to occur in collapsing supernova cores. We show how pasta phases affect the neutrino transport cross section via weak neutral current using several nuclear models. This is the first calculation of the neutrino opacity of the phases with rod-like and slab-like nuclei taking account of finite temperature effects, which are well described by the quantum molecular dynamics. We also show that pasta phases can occupy 10-20% of the mass of supernova cores in the later stage of the collapse

Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

Science.gov (United States)

Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

2004-09-01

The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

Gaseous Electron Multiplier (GEM) Detectors

Science.gov (United States)

Gnanvo, Kondo

2017-09-01

Gaseous detectors have played a pivotal role as tracking devices in the field of particle physics experiments for the last fifty years. Recent advances in photolithography and micro processing techniques have enabled the transition from Multi Wire Proportional Chambers (MWPCs) and Drift Chambers to a new family of gaseous detectors refer to as Micro Pattern Gaseous Detectors (MPGDs). MPGDs combine the basic gas amplification principle with micro-structure printed circuits to provide detectors with excellent spatial and time resolution, high rate capability, low material budget and high radiation tolerance. Gas Electron Multiplier (GEMs) is a well-established MPGD technology invented by F. Sauli at CERN in 1997 and deployed various high energy physics (HEP) and nuclear NP experiment for tracking systems of current and future NP experiments. GEM detector combines an exceptional high rate capability (1 MHz / mm2) and robustness against harsh radiation environment with excellent position and timing resolution performances. Recent breakthroughs over the past decade have allowed the possibility for large area GEMs, making them cost effective and high-performance detector candidates to play pivotal role in current and future particle physics experiments. After a brief introduction of the basic principle of GEM technology, I will give a brief overview of the GEM detectors used in particle physics experiments over the past decades and especially in the NP community at Thomas Jefferson National Laboratory (JLab) and Brookhaven National Laboratory (BNL). I will follow by a review of state of the art of the new GEM development for the next generation of colliders such as Electron Ion Collider (EIC) or High Luminosity LHC and future Nuclear Physics experiments. I will conclude with a presentation of the CERN-based RD51 collaboration established in 2008 and its major achievements regarding technological developments and applications of MPGDs.

Heterogeneous gas core reactor

International Nuclear Information System (INIS)

Han, K.I.

1977-01-01

Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

Rapid monitoring of gaseous fission products in BWRs using a portable spectrometer

International Nuclear Information System (INIS)

Yeh, Wei-Wen; Lee, Cheng-Jong; Chen, Chen-Yi; Chung, Chien

1996-01-01

Rapid, quantitative determination of gaseous radionuclides is the most difficult task in the field of environmental monitoring for radiation. Although the identification of each gaseous radionuclide is relatively straightforward using its decayed gamma ray as an index, the quantitative measurement is hampered by the time-consuming sample collection procedures, in particular for the radioactive noble gaseous fission products of krypton and xenon. In this work, a field gamma-ray spectrometer consisting of a high-purity germanium detector, portable multichannel anlayzer, and a notebook computer was used to conduct rapid scanning of radioactive krypton and xenon in the air around a nuclear facility

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Energy Technology Data Exchange (ETDEWEB)

Ehud Greenspan

2008-09-30

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Tests of Micro-Pattern Gaseous Detectors for active target time projection chambers in nuclear physics

Energy Technology Data Exchange (ETDEWEB)

Pancin, J., E-mail: pancin@ganil.fr [GANIL, CEA/DSM-CNRS/IN2P3, Bvd H. Becquerel, Caen (France); Damoy, S.; Perez Loureiro, D. [GANIL, CEA/DSM-CNRS/IN2P3, Bvd H. Becquerel, Caen (France); Chambert, V.; Dorangeville, F. [IPNO, CNRS/IN2P3, Orsay (France); Druillole, F. [CEA, DSM/Irfu/SEDI, Gif-Sur-Yvette (France); Grinyer, G.F. [GANIL, CEA/DSM-CNRS/IN2P3, Bvd H. Becquerel, Caen (France); Lermitage, A.; Maroni, A.; Noël, G. [IPNO, CNRS/IN2P3, Orsay (France); Porte, C.; Roger, T. [GANIL, CEA/DSM-CNRS/IN2P3, Bvd H. Becquerel, Caen (France); Rosier, P. [IPNO, CNRS/IN2P3, Orsay (France); Suen, L. [GANIL, CEA/DSM-CNRS/IN2P3, Bvd H. Becquerel, Caen (France)

2014-01-21

Active target detection systems, where the gas used as the detection medium is also a target for nuclear reactions, have been used for a wide variety of nuclear physics applications since the eighties. Improvements in Micro-Pattern Gaseous Detectors (MPGDs) and in micro-electronics achieved in the last decade permit the development of a new generation of active targets with higher granularity pad planes that allow spatial and time information to be determined with unprecedented accuracy. A novel active target and time projection chamber (ACTAR TPC), that will be used to study reactions and decays of exotic nuclei at facilities such as SPIRAL2, is presently under development and will be based on MPGD technology. Several MPGDs (Micromegas and Thick GEM) coupled to a 2×2 mm{sup 2} pixelated pad plane have been tested and their performances have been determined with different gases over a wide range of pressures. Of particular interest for nuclear physics experiments are the angular and energy resolutions. The angular resolution has been determined to be better than 1° FWHM for short traces of about 4 cm in length and the energy resolution deduced from the particle range was found to be better than 5% for 5.5 MeV α particles. These performances have been compared to Geant4 simulations. These experimental results validate the use of these detectors for several applications in nuclear physics.

Tests of Micro-Pattern Gaseous Detectors for active target time projection chambers in nuclear physics

International Nuclear Information System (INIS)

Pancin, J.; Damoy, S.; Perez Loureiro, D.; Chambert, V.; Dorangeville, F.; Druillole, F.; Grinyer, G.F.; Lermitage, A.; Maroni, A.; Noël, G.; Porte, C.; Roger, T.; Rosier, P.; Suen, L.

2014-01-01

Active target detection systems, where the gas used as the detection medium is also a target for nuclear reactions, have been used for a wide variety of nuclear physics applications since the eighties. Improvements in Micro-Pattern Gaseous Detectors (MPGDs) and in micro-electronics achieved in the last decade permit the development of a new generation of active targets with higher granularity pad planes that allow spatial and time information to be determined with unprecedented accuracy. A novel active target and time projection chamber (ACTAR TPC), that will be used to study reactions and decays of exotic nuclei at facilities such as SPIRAL2, is presently under development and will be based on MPGD technology. Several MPGDs (Micromegas and Thick GEM) coupled to a 2×2 mm 2 pixelated pad plane have been tested and their performances have been determined with different gases over a wide range of pressures. Of particular interest for nuclear physics experiments are the angular and energy resolutions. The angular resolution has been determined to be better than 1° FWHM for short traces of about 4 cm in length and the energy resolution deduced from the particle range was found to be better than 5% for 5.5 MeV α particles. These performances have been compared to Geant4 simulations. These experimental results validate the use of these detectors for several applications in nuclear physics

Nuclear Human Resources Development Program using Educational Core Simulator

International Nuclear Information System (INIS)

Choi, Yu Sun; Hong, Soon Kwan

2015-01-01

KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides

Nuclear Human Resources Development Program using Educational Core Simulator

Energy Technology Data Exchange (ETDEWEB)

Choi, Yu Sun; Hong, Soon Kwan [KHNP-CRI, Daejeon (Korea, Republic of)

2015-10-15

KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides.

Thermodynamic parameters and transport coefficients of the U-C-F gas mixture in the steady flow gaseous core fission reactor

International Nuclear Information System (INIS)

Berg, M.S. van den.

1995-01-01

Thermodynamic parameters and transport coefficients have been calculated for a multicomponent reacting U-C-F gas mixture in the steady flow gaseous core fission reactor. Element abundances are consistent with thermodynamic equilibrium between the gas mixture and a cooled solid graphite wall at 2500 K. Results are presented for various pressures, a fluorine potential of 5.6 and temperatures between 2500 and 7000 K. As a result of dissociation processes of uranium and carbon fluoride compounds, ''effective'' values of thermodynamic parameters and transport coefficients show anomalous behaviour with respect to so-called ''frozen'' values. The chemical reaction energy of the U-C-F gas mixture has been calculated as the driving-force behind the process of fuel redistribution to attain criticality conditions inside a functioning reactor. (author)

Nuclear design and analysis report for KALIMER breakeven core conceptual design

International Nuclear Information System (INIS)

Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

2002-04-01

During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

Behaviour of radioiodine in gaseous effluents

International Nuclear Information System (INIS)

Barry, P.J.

1968-01-01

Because of the different chemical forms in which radioiodine occurs in the gaseous state, it is important when designing efficient filters to know the chemical forms which may be present in the effluent gases when various operations are being carried out and to know the effect of different gaseous environments on the filtration efficiency. To obtain this information it is necessary to have available reliable means of characterizing different chemical forms and to sample gaseous effluents when these operations are being carried out. This paper describes the use for identifying molecular iodine of metallic screens in a multi-component sampling pack in different gaseous environments. Using multi-component sampling packs, the fractionation of iodine nuclides between different chemical forms was measured in the effluent gases escaping from an in-pile test loop in which the fuel was deliberately ruptured by restricting the flow of coolant. Sequential samples were taken for six hours after the rupture and it was possible to follow during this period the individual behaviours of 13 '1I, 133 I and 135 I. Simultaneous samples were also obtained of the noble gases in the effluent gas stream and of the iodine nuclides in the loop coolant. Similar experiments have been carried out with a view to characterizing the different chemical behaviour of radioiodine as it is released from a variety of operations in the nuclear industry including the cutting of fuel sections in metallurgical examination caves and an incinerator. (author)

Modelling of core protection and monitoring system for PWR nuclear power plant simulator

International Nuclear Information System (INIS)

Jung Kun Lee; Byoung Sung Han

1997-01-01

A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

Core access system for nuclear reactor

International Nuclear Information System (INIS)

Andrea, C.

1977-01-01

Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

Monitoring and removal of gaseous carbon-14 species

International Nuclear Information System (INIS)

Kabat, M.J.

1979-01-01

A simple and efficient method was developed for the monitoring of low level carbon-14 in nuclear power station areas and gaseous effluent. Gaseous carbon compounds (hydrocarbons and CO) are catalytically oxidized to CO 2 , which is then absorbed on solid Ca(OH) 2 at elevated temperatures. The 14 C collected is quantitatively liberated by thermal decomposition of CaCO 3 as CO 2 , which is either measured directly by flow-through detectors or absorbed in alkali hydroxide followed by liquid scintillation counting. The method can also be used for the removal of gaseous 14 C. The Ca 14 CO 3 can be immobilized in concrete for long term disposal. Ca(OH) 2 is an inexpensive absorber. It is selective for CO 2 and has high capacity and efficiency for its absorption and retention. A theoretical evaluation of thee optium conditions for CO 2 absorption and liberation is discussed and experimental investigations are described. There is good agreement between theoretical predictions and experimental findings

Safety aspects of the design of a PWR gaseous radwaste treatment system using hydrogen recombiners

International Nuclear Information System (INIS)

Glibert, R.; Nuyt, G.; Herin, S.; Fossion, P.

1978-01-01

PWR Gaseous radwaste treatment system is essential for the reduction of impact on environment of the nuclear power plants. Decay tank system has been used for the retention of the radioactive gaseous fission products generated in the primary coolant. The use of a system combining decay tanks and hydrogen recombiner units is described in this paper. Accent is put on the safety aspects of this gaseous radwaste treatment facilitystudied by BN for a Belgian Power Plant. (author)

Contribution to the study of gaseous molecular iodine washout by natural rains

International Nuclear Information System (INIS)

Fournier-Bidoz, V.

1991-01-01

This study is part of researches about nuclear accident prediction consequences on the environment. It concerns transfering of molecular gaseous iodine into liquids and especially precipitation scavenging below the cloud (washout). Bibliographic data directly concerned with this study (iodine's aqueous chemistry, aqueous to gaseous phases transfer) and also with its global frame-work (atmospheric release from a nuclear reactor in accidental situation and the behaviour of atmospheric iodine) are presented. Several experimental approaches have been performed in laboratory and on field. An aqueous to gaseous phase transfer simulator allowed us to isolate parameters involved in absorption and desorption of the halogen. Field experiments permit to quantify dry deposition on different solutions and to get a better insight of the phenomenon. Extrapolation of the whole results to precipitation scavenging of gaseous iodine I 2 by natural rains suggests that the process is an irreversible one. Washout rate values acquired during rainy experiments with molecular iodine emission or in a laboratory rainfall simulator agree with literatures data relative to irreversibility. However and even if reversibility was efficient it was not possible to clearly exhibit it according to experimental conditions. Moreover, the analytical iodine method which leads to a good experimental study has been presented

Comparative assessment of out-of-core nuclear thermionic power systems

International Nuclear Information System (INIS)

Estabrook, W.C.; Koenig, D.R.; Prickett, W.Z.

1975-01-01

The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds. (Author)

Hyper-heuristic applied to nuclear reactor core design

International Nuclear Information System (INIS)

Domingos, R P; Platt, G M

2013-01-01

The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

Rules for design of nuclear graphite core components - some considerations and approaches

International Nuclear Information System (INIS)

Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

1978-01-01

The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

Methods and techniques of nuclear in-core fuel management

International Nuclear Information System (INIS)

Jong, A.J. de.

1992-04-01

Review of methods of nuclear in-core fuel management (the minimal critical mass problem, minimal power peaking) and calculational techniques: reactorphysical calculations (point reactivity models, continuous refueling, empirical methods, depletion perturbation theory, nodal computer programs); optimization techniques (stochastic search, linear programming, heuristic parameter optimization). (orig./HP)

Gas core nuclear thermal rocket engine research and development in the former USSR

International Nuclear Information System (INIS)

Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

1992-09-01

Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR

International Nuclear Information System (INIS)

Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

1988-01-01

1 - Description of program or function: The PWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from pressurized water reactors (PWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

International Nuclear Information System (INIS)

Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

1980-01-01

Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

Fundamental design bases for independent core cooling in Swedish nuclear power reactors

International Nuclear Information System (INIS)

Jelinek, Tomas

2015-01-01

New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

In-core nuclear fuel management optimization of VVER1000 using perturbation theory

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser

2011-01-01

In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

Gaseous nebulae

International Nuclear Information System (INIS)

Williams, R.E.

1976-01-01

Gaseous nebulae are large, tenuous clouds of ionized gas that are associated with hot stars and that emit visible light because of the energy that they receive from the ultraviolet radiation of the stars. Examples include H II regions, planetary nebulae, and nova/supernova remnants. The emphasis is on the physical processes that occur in gaseous nebulae as opposed to a study of the objects themselves. The introduction discusses thermodynamic vs. steady-state equilibrium and excitation conditions in a dilute radiation field. Subsequent sections take up important atomic processes in gaseous nebulae (particle--particle collision rates, radiative interaction rates, cross sections), the ionization equilibrium (sizes of H II regions, ionization of the heavier elements), kinetic temperature and energy balance (heating of the electrons, cooling of the electrons), and the spectra of gaseous nebulae (line fluxes in nebulae). 7 figures, 5 tables

Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97.

Science.gov (United States)

Milbradt, Jens; Sonntag, Eric; Wagner, Sabrina; Strojan, Hanife; Wangen, Christina; Lenac Rovis, Tihana; Lisnic, Berislav; Jonjic, Stipan; Sticht, Heinrich; Britt, William J; Schlötzer-Schrehardt, Ursula; Marschall, Manfred

2018-01-13

The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC) that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV) capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97

Directory of Open Access Journals (Sweden)

Jens Milbradt

2018-01-01

Full Text Available The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

A new gaseous gap conductance relationship

International Nuclear Information System (INIS)

Wesley, D.A.; Yovanovich, M.M.

1986-01-01

A new relationship for predicting the gaseous gap conductance between the fuel and clad of a nuclear fuel rod is derived. This relationship is derived from purely analytical considerations and represents a departure from approaches taken in the past. A comparison between the predictions from this new relationship and experimental measurements is presented and the agreement is very good. Predictions can be generated relatively quickly with this relationship making it attractive for fuel pin analysis codes

Radiolytical oxidation of gaseous iodine by beta radiation

International Nuclear Information System (INIS)

Kaerkelae, Teemu; Auvinen, Ari; Kekki, Tommi; Kotiluoto, Petri; Lyyraenen, Jussi; Jokiniemi, Jorma

2015-01-01

Iodine is one of the most radiotoxic fission product released from fuel during a severe nuclear power plant accident. Within the containment building, iodine compounds can react e.g. on the painted surfaces and form gaseous organic iodides. In this study, it was found out that gaseous methyl iodide (CH 3 I) is oxidised when exposed to beta radiation in an oxygen containing atmosphere. As a result, nucleation of aerosol particles takes place and the formation of iodine oxide particles is suggested. These particles are highly hygroscopic. They take up water from the air humidity and iodine oxides dissolve within the droplets. In order to mitigate the possible source term, it is of interest to understand the effect of beta radiation on the speciation of iodine.

Radiolytical oxidation of gaseous iodine by beta radiation

Energy Technology Data Exchange (ETDEWEB)

Kaerkelae, Teemu; Auvinen, Ari; Kekki, Tommi; Kotiluoto, Petri; Lyyraenen, Jussi [VTT Technical Research Centre of Finland, Espoo (Finland); Jokiniemi, Jorma [VTT Technical Research Centre of Finland, Espoo (Finland); Eastern Finland Univ., Kuopio (Finland)

2015-07-01

Iodine is one of the most radiotoxic fission product released from fuel during a severe nuclear power plant accident. Within the containment building, iodine compounds can react e.g. on the painted surfaces and form gaseous organic iodides. In this study, it was found out that gaseous methyl iodide (CH{sub 3}I) is oxidised when exposed to beta radiation in an oxygen containing atmosphere. As a result, nucleation of aerosol particles takes place and the formation of iodine oxide particles is suggested. These particles are highly hygroscopic. They take up water from the air humidity and iodine oxides dissolve within the droplets. In order to mitigate the possible source term, it is of interest to understand the effect of beta radiation on the speciation of iodine.

Model, parameter and code of environmental dispersion of gaseous effluent under normal operation from nuclear power plant with 600 MWe

International Nuclear Information System (INIS)

Hu Erbang; Gao Zhanrong

1998-06-01

The model of environmental dispersion of gaseous effluence under normal operation from a nuclear power plant with 600 MWe is established to give a mathematical expression of annual mean atmospheric dispersion factor under mixing release condition based on quality assessment of radiological environment for 30 years of Chinese nuclear industry. In calculation, the impact from calm and other following factors have been taken into account: mixing layer, dry and wet deposition, radioactive decay and buildings. The doses caused from the following exposure pathways are also given by this model: external exposure from immersion cloud and ground deposition, internal exposure due to inhalation and ingestion. The code is named as ROULEA. It contains four modules, i.e. INPUT, ANRTRI, CHIQV and DOSE for calculating 4-dimension joint frequency, annual mean atmospheric dispersion factor and doses

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.

The determination of Pu-241 by liquid scintillation counting in gaseous effluents of an incineration facility, FERAB, and the Karlsruhe Nuclear Reprocessing Plant, WAK

International Nuclear Information System (INIS)

Godoy, J.M.; Schuettelkopf, H.

1983-03-01

Although the concentration of Pu-241 in nuclear fuel to be reprocessed is high, there are only few results published about the emission of Pu-241 with gaseous and liquid effluents. Nearly no information is available, too, about the environmental contamination of nuclear installations by Pu-241. Therefore a procedure was developed to measure Pu-241 by liquid scintillation counting. Sample preparation was performed by electroplating of plutonium on stainless steel planchets. To correct the selfabsorption the linear dependence of counting efficiency in the liquid scintillation counter and the resolution in the α-spectrometer was used. (orig./HP) [de

Fuel assembly and nuclear reactor core

International Nuclear Information System (INIS)

Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

1995-01-01

The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

Polarization measurement for internal polarized gaseous targets

International Nuclear Information System (INIS)

Ye Zhenyu; Ye Yunxiu; Lv Haijiang; Mao Yajun

2004-01-01

The authors present an introduction to internal polarized gaseous targets, polarization method, polarization measurement method and procedure. To get the total nuclear polarization of hydrogen atoms (including the polarization of the recombined hydrogen molecules) in the target cell, authors have measured the parameters relating to atomic polarization and polarized hydrogen atoms and molecules. The total polarization of the target during our measurement is P T =0.853 ± 0.036. (authors)

Uranium enrichment export control guide: Gaseous diffusion

Energy Technology Data Exchange (ETDEWEB)

1989-09-01

This document was prepared to serve as a guide for export control officials in their interpretation, understanding, and implementation of export laws that relate to the Zangger International Trigger List for gaseous diffusion uranium enrichment process components, equipment, and materials. Particular emphasis is focused on items that are especially designed or prepared since export controls are required for these by States that are party to the International Nuclear Nonproliferation Treaty.

Effect of spectral characterization of gaseous fuel reactors on transmutation and burning of actinides

Energy Technology Data Exchange (ETDEWEB)

Fung, C.; Anghaie, S. [Florida Univ., Wilmington, NC (United States)

2007-07-01

Gaseous Core Reactors (GCR) are fueled with stable uranium compounds in a reflected cavity. The spectral characteristics of neutrons in GCR systems could shift from one end of the spectrum to the other end by changing design parameters such as reflector material and thickness, uranium enrichment, and the average operational temperature and pressure. The rate of actinide generation, transmutation, and burnup is highly influenced by the average neutron energy in reactor core. In particular, the production rate and isotopic mix of plutonium are highly dependent on the neutron spectrum in the reactor. Other actinides of primary interest to this work are neptunium-237 and americium-241 due to their pivotal impact on high-level nuclear waste disposal. In all cavity reactors including GCR's, the reflector material and thickness are the most important design parameters in determining the core spectrum. The increase in the gaseous fuel pressure and enrichment results in relative shift of neutron population toward energies greater than 2 eV. Reflector materials considered in this study are beryllium oxide, lithium hydride, lithium deuteride, zirconium carbide, graphite, lead, and tungsten. Results of the study suggest that the beryllium oxide and tungsten reflected GCR systems set the lower (softest) and upper (hardest) limits of neutron spectra, respectively. The inventory of actinides with half-lives greater than 1000 years can be minimized by increasing neutron flux level in the reactor core. The higher the neutron flux, the lower the inventory of these actinides. The majority of the GCR designs maintained a flux level on the order of 10{sup 15} cm{sup -2}*s{sup -1} while the PWR flux is one order of magnitude lower. The inventory of the feeder isotopes to Np{sup 237} including U{sup 237}, Pu{sup 241}, and Am{sup 241} decreases with relative shift of neutron spectrum toward higher energies. This is due to increased resonance absorption in these isotopes due to higher

Device for measuring flow rate in a nuclear reactor core

International Nuclear Information System (INIS)

Hamano, Jiro.

1980-01-01

Purpose: To always calculate core flow rate automatically and accurately in BWR type nuclear power plants. Constitution: Jet pumps are provided to the recycling pump and to the inside of the pressure vessel of a nuclear reactor. The jet pumps comprise a plurality of calibrated jet pumps for forcively convecting the coolants and a plurality of not calibrated jet pumps in order to cool the heat generated in the reactor core. The difference in the pressures between the upper and the lower portions in both of the jet pumps is measured by difference pressure transducers. Further, a thermo-sensitive element is provided to measure the temperature of recycling water at the inlet of the recycling pump. The output signal from the difference pressure transducer is inputted to a process computer, calculated periodically based on predetermined calculation equations, compensated for the temperature by a recycling water temperature signal and outputted as a core flow rate signal to a recoder. The signal is also used for the power distribution calculation in the process computer and the minimum limit power ratio as the thermal limit value for the fuels is outputted. (Furukawa, Y.)

The development of direct core monitoring in Nuclear Electric plc

International Nuclear Information System (INIS)

Curtis, R.F.; Jones, S. Reed, J.; Wickham, A.J.

1996-01-01

Monitoring of graphite behaviour in Nuclear Electric Magnox and AGR reactors is necessary to support operating safety cases and to ensure that reactor operation is optimized to sustain the necessary core integrity for the economic life of the reactors. The monitoring programme combines studies for pre-characterized ''installed'' samples with studies on samples trepanned from within the cores and also with studies of core and channel geometry using specially designed equipment. Nuclear Electric has two trepanning machines originally designed for Magnox-reactor work which have been used for a substantial programme over many years. They have recently been upgraded to improve sampling speed, safety and versatility - the last being demonstrated by their adaptation for a recently-won contract associated with decommissioning the Windscale piles. Radiological hazards perceived when the AGR trepanning system was designed resulted in very cumbersome equipment. This has worked well but has been inconvenient in operation. The development of a smaller and improved system for deploying the equipment is now reported. Channel dimension monitoring equipment is discussed in detail with examples of data recovered from both Magnox and AGR cores. A resolution of ± 2 of arc (tilt) and ± 0.01 mm change in diameter in attainable. It is also theoretically possible to establish brick stresses by measuring geometry changes which result from trepanning. Current development work on a revolving scanning laser rangefinder which will enable the measurement of diameters to a resolution of 0.001 mm will also be discussed. This paper also discusses non-destructive techniques for crack detection employing ultrasound or resistance networks, the use of special manipulators to deliver inspection and repair equipment and recent developments to install displacement monitors in peripheral regions of the cores, to aid the understanding of the interaction of the restraint system with the core - the region

Uranium enrichment by the gaseous diffusion process

International Nuclear Information System (INIS)

Petit, J.F.

1977-01-01

After a brief description of the process and technology (principle, stage constitution, cascade constitution, and description of a plant), the author gives the history of gaseous diffusion and describes the existing facilities. Among the different enrichment processes contemplated in the USA during and after the last world war, gaseous diffusion has been the only one to be developed industrially on a wide scale in the USA, then in the UK, in the USSR and in France. The large existing capacities in the USA provided the country with a good starting base for the development of a light-water nuclear power plant programme, the success of which led to a shortfall in production means. France and the USA, possessing the necessary know-how, have been able to undertake the realization of two industrial programmes: the CIP-CUP programme in the USA and the Eurodif project in France. Current plans still call for the construction of several plants by 1990. Can the gaseous diffusion process meet this challenge. Technically, there is no doubt about it. Economically, this process is mainly characterized by large energy consumption and the necessity to build large plants. This leads to a large investment, at least for the first plant. Other processes have been developed with a view to reducing both energy and capital needs. However, in spite of continuous studies and technological progress, no process has yet proved competitive. Large increments in capacities are still expected to come from gaseous diffusion, and several projects taking into account the improvements in flexibility, automatization, reliability and reduced investment, are analysed in the paper. Combining new facilities with existing plants has already proved to be of great interest. This situation explains why gaseous diffusion is being further investigated and new processes are being studied. (author)

Gaseous swelling of B4C and UO2 fuel: similarities and differences

International Nuclear Information System (INIS)

Evdokimov, I.; Khoruzhii, O.; Kourtchatov, S.; Likhanskii, V.; Matweev, L.

2001-01-01

A major factor limiting the resource of control rods (CRs) for WWER-1000 reactors is their radiation damage. Radiation induced embrittlement of the CRs cladding, core swelling and gaseous internal pressure in CRs result in mechanical core-cladding interaction. This work is devoted to the physical analysis of processes that control the structural changes in neutron absorber elements with B 4 C under irradiation in water reactors. Particularly, the analysis of mechanisms of the helium porosity formation in B 4 C is undertaken. In view of the deficiency of experimental data on the subject, a fruitful approach to the problem is a comparative analysis of the swelling mechanisms in B 4 C absorber and UO 2 fuel. Using this similarity a phenomenological model of fission gas behavior in boron carbide is proposed. The model predictions for radial profile of 10 B burnup under influence of thermal and epithermal neutrons are compared with experimental results. The main results show that despite the external similarity of the process of fission gas accumulation in UO 2 and in B 4 C, phenomenology of gaseous swelling is much different for the fuel and the CR core. The reason for that difference is the distinction of physical conditions in irradiated fuel and CR core

A new equation of state for core-collapse supernovae based on realistic nuclear forces and including a full nuclear ensemble

International Nuclear Information System (INIS)

Furusawa, S; Togashi, H; Nagakura, H; Sumiyoshi, K; Yamada, S; Suzuki, H; Takano, M

2017-01-01

We have constructed a nuclear equation of state (EOS) that includes a full nuclear ensemble for use in core-collapse supernova simulations. It is based on the EOS for uniform nuclear matter that two of the authors derived recently, applying a variational method to realistic two- and three-body nuclear forces. We have extended the liquid drop model of heavy nuclei, utilizing the mass formula that accounts for the dependences of bulk, surface, Coulomb and shell energies on density and/or temperature. As for light nuclei, we employ a quantum-theoretical mass evaluation, which incorporates the Pauli- and self-energy shifts. In addition to realistic nuclear forces, the inclusion of in-medium effects on the full ensemble of nuclei makes the new EOS one of the most realistic EOSs, which covers a wide range of density, temperature and proton fraction that supernova simulations normally encounter. We make comparisons with the FYSS EOS, which is based on the same formulation for the nuclear ensemble but adopts the relativistic mean field theory with the TM1 parameter set for uniform nuclear matter. The new EOS is softer than the FYSS EOS around and above nuclear saturation densities. We find that neutron-rich nuclei with small mass numbers are more abundant in the new EOS than in the FYSS EOS because of the larger saturation densities and smaller symmetry energy of nuclei in the former. We apply the two EOSs to 1D supernova simulations and find that the new EOS gives lower electron fractions and higher temperatures in the collapse phase owing to the smaller symmetry energy. As a result, the inner core has smaller masses for the new EOS. It is more compact, on the other hand, due to the softness of the new EOS and bounces at higher densities. It turns out that the shock wave generated by core bounce is a bit stronger initially in the simulation with the new EOS. The ensuing outward propagations of the shock wave in the outer core are very similar in the two simulations, which

Methods to prevent the source term of methyl lodide during a core melt accident

Energy Technology Data Exchange (ETDEWEB)

Karhu, A. [VTT Energy (Finland)

1999-11-01

The purpose of this literature review is to gather available information of the methods to prevent a source term of methyl iodide during a core melt accident. The most widely studied methods for nuclear power plants include the impregnated carbon filters and alkaline additives and sprays. It is indicated that some deficiencies of these methods may emerge. More reactive impregnants and additives could make a great improvement. As a new method in the field of nuclear applications, the potential of transition metals to decompose methyl iodide, is introduced in this review. This area would require an additional research, which could elucidate the remaining questions of the reactions. The ionization of the gaseous methyl iodide by corona-discharge reactors is also shortly described. (au)

Nuclear cardiology core syllabus of the European Association of Cardiovascular Imaging (EACVI).

Science.gov (United States)

Gimelli, Alessia; Neglia, Danilo; Schindler, Thomas H; Cosyns, Bernard; Lancellotti, Patrizio; Kitsiou, Anastasia

2015-04-01

The European Association of Cardiovascular Imaging (EACVI) Core Syllabus for Nuclear Cardiology is now available online. The syllabus lists key elements of knowledge in nuclear cardiology. It represents a framework for the development of training curricula and provides expected knowledge-based learning outcomes to the nuclear cardiology trainees. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2015. For permissions please email: journals.permissions@oup.com.

Nuclear criticality safety controls for uranium deposits during D and D at the Oak Ridge Gaseous Diffusion Plant

International Nuclear Information System (INIS)

Haire, M.J.; Jordan, W.C.; Jollay, L.J. III; Dahl, T.L.

1997-01-01

The US Department of Energy (DOE) Deputy Assistant Secretary of Energy for Environmental Management has issued a challenge to complete DOE environmental cleanup within a decade. The response for Oak Ridge facilities is in accordance with the DOE ten-year plan which calls for completion of > 95% of environmental management work by the year 2006. This will result in a 99% risk reduction and in a significant savings in base line costs in waste management (legacy waste); remedial action (groundwater, soil, etc.); and decontamination and decommissioning (D and D). It is assumed that there will be long-term institutional control of cascade equipment, i.e., there will be no walk away from sites, and that there will be firm radioactivity release limits by 1999 for recycle metals. An integral part of these plants is the removal of uranium deposits which pose nuclear criticality safety concerns in the shut down of the Oak Ridge Gaseous Diffusion Plant. DOE has initiated the Nuclear Criticality Stabilization Program to improve nuclear criticality safety by removing the larger uranium deposits from unfavorable geometry equipment. Nondestructive assay (NDA) measurements have identified the location of these deposits. The objective of the K-25 Site Nuclear Criticality Stabilization Program is to remove and place uranium deposits into safe geometry storage containers to meet the double contingency principle. Each step of the removal process results in safer conditions where multiple controls are present. Upon completion of the Program, nuclear criticality risks will be greatly reduced

Analysis of ex-core detector response measured during nuclear ship Mutsu land-loaded core critical experiment

International Nuclear Information System (INIS)

Itagaki, M.; Abe, J.I.; Kuribayashi, K.

1987-01-01

There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment

The Impact of the Nuclear Equation of State in Core Collapse Supernovae

Science.gov (United States)

Baird, M. L.; Lentz, E. J.; Hix, W. R.; Mezzacappa, A.; Messer, O. E. B.; Liebendoerfer, M.; TeraScale Supernova Initiative Collaboration

2005-12-01

One of the key ingredients to the core collapse supernova mechanism is the physics of matter at or near nuclear density. Included in simulations as part of the Equation of State (EOS), nuclear repulsion experienced at high densities are responsible for the bounce shock, which initially causes the outer envelope of the supernova to expand, as well as determining the structure of the newly formed proto-neutron star. Recent years have seen renewed interest in this fundamental piece of supernova physics, resulting in several promising candidate EOS parameterizations. We will present the impact of these variations in the nuclear EOS using spherically symmetric, Newtonian and General Relativistic neutrino transport simulations of stellar core collapse and bounce. This work is supported in part by SciDAC grants to the TeraScale Supernovae Initiative from the DOE Office of Science High Energy, Nuclear, and Advanced Scientific Computing Research Programs. Oak Ridge National Laboratory is managed by UT-Battelle, LLC, for U.S. Department of Energy under contract DEAC05-00OR22725

Enhanced core monitoring system for Browns Ferry Nuclear Plant

International Nuclear Information System (INIS)

Lindsey, R.S.

1980-01-01

A system of computer hardware and software is being developed to supplement the process computers at Browns Ferry Nuclear Plant in the area of reactor core monitoring. All data stored in the process computers will be made available through a data link to an onsite minicomputer which will store and edit the data for engineering and operations personnel. Important core parameters will be effectively displayed on color graphic CRT terminals using techniques such as blinking, shading, and color coding to emphasize significant values. This data will also be made available to Tennessee Valley Authority's Chattanooga central office support groups through a data network between existing computers

Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

International Nuclear Information System (INIS)

Ehud Greenspan

2008-01-01

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

International Nuclear Information System (INIS)

Prelas, M.A.; Boody, F.P.; Zediker, M.S.

1987-01-01

An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

The effects of radiation on aluminium alloys in the core of energy nuclear reactors

International Nuclear Information System (INIS)

Petrossian, V.G.

1995-01-01

One of the attractive directions in the worldwide practice of nuclear installations is the replacement of expensive zirconium alloy with more cheap materials, particularly aluminium allo. For Heat Supply Nuclear Plants (HSNP) with approximately 473 K core temperatures, the use of heat-resistant aluminium alloys seems to be reasonable. The present work is concerned with the studies on radiation effects on aluminium alloy, and interaction between the alloy and coolant in the reactor core. (author). 2 refs., 3 figs., 1 tab

Optimization analysis of the nuclear fuel cycle transition to the last core

International Nuclear Information System (INIS)

Rebollo, L.; Blanco, J.

2001-01-01

The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235 U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235 U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

Studies on the inhomogeneous core density of a fluidized bed nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Van der Hagen, T.H.J.J.; Van Dam, H.; Hoogenboom, J.E.; Khotylev, V.A. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.; Harteveld, W.; Mudde, R.F.

1997-12-31

Results are reported on the expected time dependent core density profile of a fluidized-bed nuclear fission reactor. Core densities have been measured in a test facility by the gamma-transmission technique. Bubble and particle-cluster sizes, positions, velocities and frequencies could be determined. Neutronic studies have been performed on the influence of core voids on reactivity using Monte-Carlo and neutron-transport codes. Fuel-particle importance has been determined. Point-kinetic parameters have been calculated for linking reactivity perturbations to power fluctuations. (author)

Mass-radius relations and core-envelope decompositions of super-Earths and sub-Neptunes

Energy Technology Data Exchange (ETDEWEB)

Howe, Alex R.; Burrows, Adam [Department of Astrophysical Sciences, Princeton University, Peyton Hall, Princeton, NJ 08544 (United States); Verne, Wesley, E-mail: arhowe@astro.princeton.edu, E-mail: burrows@astro.princeton.edu [Department of Computer Science, Princeton University, Princeton, NJ 08544 (United States)

2014-06-01

Many exoplanets have been discovered with radii of 1-4 R {sub ⊕}, between that of Earth and Neptune. A number of these are known to have densities consistent with solid compositions, while others are 'sub-Neptunes' likely to have significant H{sub 2}-He envelopes. Future surveys will no doubt significantly expand these populations. In order to understand how the measured masses and radii of such planets can inform their structures and compositions, we construct models both for solid layered planets and for planets with solid cores and gaseous envelopes, exploring a range of core masses, H{sub 2}-He envelope masses, and associated envelope entropies. For planets in the super-Earth/sub-Neptune regime for which both radius and mass are measured, we estimate how each is partitioned into a solid core and gaseous envelope, associating a specific core mass and envelope mass with a given exoplanet. We perform this decomposition for both ''Earth-like'' rock-iron cores and pure ice cores, and find that the necessary gaseous envelope masses for this important sub-class of exoplanets must range very widely from zero to many Earth masses, even for a given core mass. This result bears importantly on exoplanet formation and envelope evaporation processes.

Mass-radius relations and core-envelope decompositions of super-Earths and sub-Neptunes

International Nuclear Information System (INIS)

Howe, Alex R.; Burrows, Adam; Verne, Wesley

2014-01-01

Many exoplanets have been discovered with radii of 1-4 R ⊕ , between that of Earth and Neptune. A number of these are known to have densities consistent with solid compositions, while others are 'sub-Neptunes' likely to have significant H 2 -He envelopes. Future surveys will no doubt significantly expand these populations. In order to understand how the measured masses and radii of such planets can inform their structures and compositions, we construct models both for solid layered planets and for planets with solid cores and gaseous envelopes, exploring a range of core masses, H 2 -He envelope masses, and associated envelope entropies. For planets in the super-Earth/sub-Neptune regime for which both radius and mass are measured, we estimate how each is partitioned into a solid core and gaseous envelope, associating a specific core mass and envelope mass with a given exoplanet. We perform this decomposition for both ''Earth-like'' rock-iron cores and pure ice cores, and find that the necessary gaseous envelope masses for this important sub-class of exoplanets must range very widely from zero to many Earth masses, even for a given core mass. This result bears importantly on exoplanet formation and envelope evaporation processes.

Core Technology Development of Nuclear spin polarization

International Nuclear Information System (INIS)

Yoo, Byung Duk; Gwon, Sung Ok; Kwon, Duck Hee; Lee, Sung Man

2009-12-01

In order to study nuclear spin polarization, we need several core technologies such as laser beam source to polarize the nuclear spin, low pressured helium cell development whose surface is essential to maintain polarization otherwise most of the polarized helium relaxed in short time, development of uniform magnetic field system which is essential for reducing relaxation, efficient vacuum system, development of polarization measuring system, and development of pressure raising system about 1000 times. The purpose of this study is to develop resonable power of laser system, that is at least 5 watt, 1083 nm, 4GHz tuneable. But the limitation of this research fund enforce to develop amplifying system into 5 watt with 1 watt system utilizing laser-diod which is already we have in stock. We succeeded in getting excellent specification of fiber laser system with power of 5 watts, 2 GHz linewidth, more than 80 GHz tuneable

Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

International Nuclear Information System (INIS)

Gehre, G.

1982-01-01

First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

International Nuclear Information System (INIS)

Castiblanco, L.A.; Sarta, J.A.

1997-01-01

The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

Automated software analysis of nuclear core discharge data

International Nuclear Information System (INIS)

Larson, T.W.; Halbig, J.K.; Howell, J.A.; Eccleston, G.W.; Klosterbuer, S.F.

1993-03-01

Monitoring the fueling process of an on-load nuclear reactor is a full-time job for nuclear safeguarding agencies. Nuclear core discharge monitors (CDMS) can provide continuous, unattended recording of the reactor's fueling activity for later, qualitative review by a safeguards inspector. A quantitative analysis of this collected data could prove to be a great asset to inspectors because more information can be extracted from the data and the analysis time can be reduced considerably. This paper presents a prototype for an automated software analysis system capable of identifying when fuel bundle pushes occurred and monitoring the power level of the reactor. Neural network models were developed for calculating the region on the reactor face from which the fuel was discharged and predicting the burnup. These models were created and tested using actual data collected from a CDM system at an on-load reactor facility. Collectively, these automated quantitative analysis programs could help safeguarding agencies to gain a better perspective on the complete picture of the fueling activity of an on-load nuclear reactor. This type of system can provide a cost-effective solution for automated monitoring of on-load reactors significantly reducing time and effort

Development of gaseous photomultiplier

International Nuclear Information System (INIS)

Tokanai, F.; Sumiyoshi, T.; Sugiyama, H.; Okada, T.

2014-01-01

We have been developing gaseous photomultiplier tubes (PMTs) with alkali photocathode combined with micropattern gas detectors (MPGDs). The potential advantage of the gaseous PMT is that it can achieve a very large effective area with adequate position and timing resolutions. In addition, it will be easily operated under a very high magnetic field, compared with the conventional vacuum-based PMT. To evaluate the gaseous PMTs filled with Ne and Ar based gas mixture, we have developed gaseous PMTs with an alkali photocathode combined with MPGDs such as a glass capillary plate, GEM, and Micromegas detector. We describe the recent development of the gaseous PMTs, particularly the production of the photocathode, gas gain, ion and photon feedbacks, quantum efficiency, and the characteristics in the magnetic field environment. (author)

An assessment of coupling algorithms for nuclear reactor core physics simulations

Energy Technology Data Exchange (ETDEWEB)

Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

2016-04-15

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

Nuclear equation of state for core-collapse supernova simulations with realistic nuclear forces

Energy Technology Data Exchange (ETDEWEB)

Togashi, H., E-mail: hajime.togashi@riken.jp [Nishina Center for Accelerator-Based Science, Institute of Physical and Chemical Research (RIKEN), 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Nakazato, K. [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Takehara, Y.; Yamamuro, S.; Suzuki, H. [Department of Physics, Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan); Takano, M. [Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Department of Pure and Applied Physics, Graduate School of Advanced Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan)

2017-05-15

A new table of the nuclear equation of state (EOS) based on realistic nuclear potentials is constructed for core-collapse supernova numerical simulations. Adopting the EOS of uniform nuclear matter constructed by two of the present authors with the cluster variational method starting from the Argonne v18 and Urbana IX nuclear potentials, the Thomas–Fermi calculation is performed to obtain the minimized free energy of a Wigner–Seitz cell in non-uniform nuclear matter. As a preparation for the Thomas–Fermi calculation, the EOS of uniform nuclear matter is modified so as to remove the effects of deuteron cluster formation in uniform matter at low densities. Mixing of alpha particles is also taken into account following the procedure used by Shen et al. (1998, 2011). The critical densities with respect to the phase transition from non-uniform to uniform phase with the present EOS are slightly higher than those with the Shen EOS at small proton fractions. The critical temperature with respect to the liquid–gas phase transition decreases with the proton fraction in a more gradual manner than in the Shen EOS. Furthermore, the mass and proton numbers of nuclides appearing in non-uniform nuclear matter with small proton fractions are larger than those of the Shen EOS. These results are consequences of the fact that the density derivative coefficient of the symmetry energy of our EOS is smaller than that of the Shen EOS.

Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

International Nuclear Information System (INIS)

Ishikawa, Makoto

1994-01-01

In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

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

1978-01-01

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

Dispositions taken in France to limit gaseous releases from PWR power plants in abnormal operating conditions

International Nuclear Information System (INIS)

Collinet, J.; Guieu, S.; Mulcey, P.

1989-12-01

The implementation of France's major nuclear programme - 56 PWR units in service or under construction - has gone hand in hand with the development of an original philosophy in the field of nuclear safety. From an initial core of deterministic safety philosophy current in the seventies, which has been wholly retained and, in some instances, refined, a range of complements has been made to include consideration of a number of additional situations based on a probabilistic approach. This has resulted in a better coherence for safety and a reduction of the severe accident probability. Furthermore, the establishment of emergency plans has enabled the Safety Authorities and the utility to adopt a coherent and logical approach to severe accidents, with the aim of better achieving defence in depth. This has resulted in the provision of certain additional measures intended to further reduce the consequences of severe accidents. In a accordance with the safety philosophy, adopted in France for nuclear PWR power stations, filtration systems have been specified and installed to limit the radiological consequences of consecutive gaseous emissions, on the one hand, in accidents taken into account in the design and, on the other hand, in accidents liable to jeopardize the integrity of the containment

Decision no 2009-DC-0157 of the 15. of September 2009 by the Nuclear Safety Authority specifying the limits of releases in the environment of liquid and gaseous effluents of the base nuclear installation n. 29 operated by the CIS Bio International on the district of Saclay (Essonne department)

International Nuclear Information System (INIS)

2009-01-01

This document contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents of a base nuclear installation. Tables present the limits for radioactive liquid effluents (iodine, rare earths, and other beta and gamma emitters) and the maximum admitted concentrations for gaseous chemical effluents (SO 2 , H 2 S, VOCs)

Core of a fast neutron nuclear reactor

International Nuclear Information System (INIS)

Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

1974-01-01

The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

Sensors for use in nuclear reactor cores

International Nuclear Information System (INIS)

1980-01-01

A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter. (UK)

Sensors for use in nuclear reactor cores

Energy Technology Data Exchange (ETDEWEB)

1980-05-21

A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter.

EURODIF: the uranium enrichment by gaseous diffusion

International Nuclear Information System (INIS)

Rougeau, J.P.

1981-01-01

During the seventies the nuclear power programme had an extremely rapid growth rate which entailed to increase the world uranium enrichment capacity. EURODIF is the largest undertaking in this field. This multinational joint venture built and now operates and enrichment plant using the gaseous diffusion process at Tricastin (France). This plant is delivering low enriched uranium since two years and has contracted about 110 million SWU's till 1990. Description, current activity and prospects are given in the paper. (Author) [pt

Westinghouse Nuclear Core Design Training Center - a design simulator

International Nuclear Information System (INIS)

Altomare, S.; Pritchett, J.; Altman, D.

1992-01-01

The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks; Recarga de reatores nucleares utilizando redes conectivas de colonias de formigas artificiais

Energy Technology Data Exchange (ETDEWEB)

Lima, Alan M.M. de; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: alan@lmp.ufrj.br; schirru@lmp.ufrj.br

2005-07-01

A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 2. BFS-62 analysis results

International Nuclear Information System (INIS)

Mantourov, Guennadi

2001-11-01

This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. Data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS-62 and ZPPR JUPIER series fast reactor cores, applying JNC core calculation code CITATION-FBR. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the First Flight Collision Probability method and subgroup approach. Especially, two converting programs were written to transmit the prepared effective cross sections to JNC standard PDS files to let then the CITATION code be applied for 3-D HEXZ neutronics calculations of the investigated cores. The effects of nuclear data library have been studied by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for 4 BFS-62 cores as well as for 3 ZPPR JUPITER experiment series cores ZPPR-9, ZPPR-13A and ZPPR-17A are presented. The comparison results for reaction rates distributions for 2 BFS-62 uranium loaded cores are included too. The calculated correction factors applied in all cases were less than 1.0%. The estimated uncertainty in k-effective C values caused by possible errors in calculation of the applied corrections is about 0.3% in case of BFS-62 and ZPPR MOX cores, and is about 0.2% for BFS-62 uranium-loaded cores. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC's calculation route for k-effective results is about 0.3% for ZPPR and BFS-62 cores with plutonium. As for BFS uranium-loaded cores (BFS-62-1 and BFS-62-2) the nuclear data library effect is about 0.1%. Next the sensitivity analysis was applied. It shown that the main contributors to the nuclear data library effect

A system for obtaining an optimized pre design of nuclear reactor core

International Nuclear Information System (INIS)

Mai, L.A.

1989-01-01

This work proposes a method for obtaing a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one-energy-group, unidimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, refletor thickness, enrichement and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for futures works. (author) [pt

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

International Nuclear Information System (INIS)

Clark, J.S.; Walton, J.T.; Mcguire, M.L.

1992-07-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

Radioiodine removal in nuclear facilities

International Nuclear Information System (INIS)

1980-01-01

Technical means are reviewed available for the retention of radioiodine in nuclear power plants and fuel reprocessing plants, its immobilization, storage, and disposal. The removal of iodine species from gaseous effluents of nuclear power plants using impregnated activated charcoal is dealt with. Various scrubbing techniques for trapping iodine from the head-end and dissolver off-gases are discussed as well as solid adsorbents for iodine which may be used to clean up other gaseous streams. Current practices and activities for radioiodine treatment and management in Belgian, Dutch, Swedish, USSR and UK nuclear installations are presented

Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

International Nuclear Information System (INIS)

Mantourov, Guennadi

2001-05-01

This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

Neutrino-pair emission from nuclear de-excitation in core-collapse supernova simulations

Science.gov (United States)

Fischer, T.; Langanke, K.; Martínez-Pinedo, G.

2013-12-01

We study the impact of neutrino-pair production from the de-excitation of highly excited heavy nuclei on core-collapse supernova simulations, following the evolution up to several 100 ms after core bounce. Our study is based on the agile-boltztransupernova code, which features general relativistic radiation hydrodynamics and accurate three-flavor Boltzmann neutrino transport in spherical symmetry. In our simulations the nuclear de-excitation process is described in two different ways. At first we follow the approach proposed by Fuller and Meyer [Astrophys. J.AJLEEY0004-637X10.1086/170317 376, 701 (1991)], which is based on strength functions derived in the framework of the nuclear Fermi-gas model of noninteracting nucleons. Second, we parametrize the allowed and forbidden strength distributions in accordance with measurements for selected nuclear ground states. We determine the de-excitation strength by applying the Brink hypothesis and detailed balance. For both approaches, we find that nuclear de-excitation has no effect on the supernova dynamics. However, we find that nuclear de-excitation is the leading source for the production of electron antineutrinos as well as heavy-lepton-flavor (anti)neutrinos during the collapse phase. At sufficiently high densities, the associated neutrino spectra are influenced by interactions with the surrounding matter, making proper simulations of neutrino transport important for the determination of the neutrino-energy loss rate. We find that, even including nuclear de-excitations, the energy loss during the collapse phase is overwhelmingly dominated by electron neutrinos produced by electron capture.

Comment on the in-core measurement in the WWER nuclear power plant

International Nuclear Information System (INIS)

Krett, V.; Dach, K.; Erben, O.

1985-01-01

The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

Core power distribution measurement and data processing in Daya Bay Nuclear Power Station

International Nuclear Information System (INIS)

Zhang Hong

1997-01-01

For the first time in China, Daya Bay Nuclear Power Station applied the advanced technology of worldwide commercial pressurized reactors to the in-core detectors, the leading excore six-chamber instrumentation for precise axial power distribution, and the related data processing. Described in this article are the neutron flux measurement in Daya Bay Nuclear Power Station, and the detailed data processing

Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

International Nuclear Information System (INIS)

Vondy, D.R.

1986-04-01

Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

Development of Core Monitoring System for Nuclear Power Plants (I)

Energy Technology Data Exchange (ETDEWEB)

Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

1997-12-31

1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

Feasibility study on nuclear core design for soluble boron free small modular reactor

Energy Technology Data Exchange (ETDEWEB)

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-04-29

A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

Science.gov (United States)

Kazeminezhad, F.; Anghaie, S.

2008-01-01

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

Protein kinases responsible for the phosphorylation of the nuclear egress core complex of human cytomegalovirus.

Science.gov (United States)

Sonntag, Eric; Milbradt, Jens; Svrlanska, Adriana; Strojan, Hanife; Häge, Sigrun; Kraut, Alexandra; Hesse, Anne-Marie; Amin, Bushra; Sonnewald, Uwe; Couté, Yohann; Marschall, Manfred

2017-10-01

Nuclear egress of herpesvirus capsids is mediated by a multi-component nuclear egress complex (NEC) assembled by a heterodimer of two essential viral core egress proteins. In the case of human cytomegalovirus (HCMV), this core NEC is defined by the interaction between the membrane-anchored pUL50 and its nuclear cofactor, pUL53. NEC protein phosphorylation is considered to be an important regulatory step, so this study focused on the respective role of viral and cellular protein kinases. Multiply phosphorylated pUL50 varieties were detected by Western blot and Phos-tag analyses as resulting from both viral and cellular kinase activities. In vitro kinase analyses demonstrated that pUL50 is a substrate of both PKCα and CDK1, while pUL53 can also be moderately phosphorylated by CDK1. The use of kinase inhibitors further illustrated the importance of distinct kinases for core NEC phosphorylation. Importantly, mass spectrometry-based proteomic analyses identified five major and nine minor sites of pUL50 phosphorylation. The functional relevance of core NEC phosphorylation was confirmed by various experimental settings, including kinase knock-down/knock-out and confocal imaging, in which it was found that (i) HCMV core NEC proteins are not phosphorylated solely by viral pUL97, but also by cellular kinases; (ii) both PKC and CDK1 phosphorylation are detectable for pUL50; (iii) no impact of PKC phosphorylation on NEC functionality has been identified so far; (iv) nonetheless, CDK1-specific phosphorylation appears to be required for functional core NEC interaction. In summary, our findings provide the first evidence that the HCMV core NEC is phosphorylated by cellular kinases, and that the complex pattern of NEC phosphorylation has functional relevance.

The density jump at the inner core boundary using underground nuclear explosion records

International Nuclear Information System (INIS)

Krasnoshchekov, D.N.; Ovchinnikov, V.M.

2001-01-01

This paper presents the estimation of the minimum jump value using experimental wave forms reflected from the boundary between the Earth core and mantle (PcP) and the one between the inner and outer core (PKiKP) at a distance of 6 deg. Digital seismic records of underground nuclear tests conducted at the Semipalatinsk test site in 70s by Zerenda-Vostochny-Chkalovo seismic array have been used. (author)

NUCORE - A system for nuclear structure calculations with cluster-core models

International Nuclear Information System (INIS)

Heras, C.A.; Abecasis, S.M.

1982-01-01

Calculation of nuclear energy levels and their electromagnetic properties, modelling the nucleus as a cluster of a few particles and/or holes interacting with a core which in turn is modelled as a quadrupole vibrator (cluster-phonon model). The members of the cluster interact via quadrupole-quadrupole and pairing forces. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

In core instrumentation for online nuclear heating measurements of material testing reactor

International Nuclear Information System (INIS)

Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

2010-01-01

The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

Analytical measurements of fission products during a severe nuclear accident

Science.gov (United States)

Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

2018-01-01

The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

Analytical measurements of fission products during a severe nuclear accident

Directory of Open Access Journals (Sweden)

Doizi D.

2018-01-01

Full Text Available The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

2014-06-15

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (k{sub eff}), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

Computation system for nuclear reactor core analysis

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

1977-04-01

This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

Nuclear-station post-accident liquid-sampling system: developed by Duke Power Company

International Nuclear Information System (INIS)

Burton, D.A.; Birch, M.L.; Orth, W.C.

1981-01-01

The accident at Three Mile Island showed that means must be provided to determine the radioactivity levels in high activity liquid and gaseous systems of a nuclear power plant without undue radiation exposure to personnel. The Duke Power Post Accident Liquid Sampling System provides the means for obtaining diluted liquid samples and diluted dissolved gas samples following a reactor accident involving substantial core damage. Their approach yields a straightforward engineering solution at a fraction of the cost of other systems. A description of the system, general design criteria, and color coded flow diagrams are included

Gaseous Matter

CERN Document Server

Angelo, Joseph A

2011-01-01

aseous Matter focuses on the many important discoveries that led to the scientific interpretation of matter in the gaseous state. This new, full-color resource describes the basic characteristics and properties of several important gases, including air, hydrogen, helium, oxygen, and nitrogen. The nature and scope of the science of fluids is discussed in great detail, highlighting the most important scientific principles upon which the field is based. Chapters include:. Gaseous Matter An Initial Perspective. Physical Characteristics of Gases. The Rise of the Science of Gases. Kinetic Theory of

A system to obtain an optimized first design of a nuclear reactor core

International Nuclear Information System (INIS)

Mai, L.A.

1988-01-01

This work proposes a method for obtaining a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one energy-group, one-dimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, reflector thickness, enrichment and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for future works. (autor)

Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2011-01-01

Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

Methodology for assessment of safety risk due to potential accidents in US gaseous diffusion plants

International Nuclear Information System (INIS)

Turner, J.H.; O'Kain, D.U.

1991-01-01

Gaseous diffusion plants that operate in the United States represent a unique combination of nuclear and chemical hazards. Assessing and controlling the health, safety, and environmental risks that can result from natural phenomena events, process upset conditions, and operator errors require a unique methodology. Such a methodology has been developed for the diffusion plants and is being utilized to assess and control the risk of operating the plants. A summary of the methodology developed to assess the unique safety risks at the US gaseous diffusion plants is presented in this paper

The Blend Down Monitoring System Demonstration at the Padijcah Gaseous Diffusion Plant

International Nuclear Information System (INIS)

Benton, J.; Close, D.; Johnson, W. Jr.; Kerr, P.; March-Leuba, J.; Mastal, E.; Moss, C.; Powell, D.; Sumner, J.; Uckan, T.; Vines, R.; Wright, P.D.

1999-01-01

Agreements between the governments of the US and the Russian Federation for the US purchase of low enriched uranium (LEU) derived from highly enriched uranium (HEU) from dismantled Russian nuclear weapons calls for the establishment of transparency measures to provide confidence that nuclear nonproliferation goals are being met. To meet these transparency goals, the agreements call for the installation of nonintrusive US instruments to monitor the down blending of HEU to LEU. The Blend Down Monitoring System (BDMS) has been jointly developed by the Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) to continuously monitor 235 U enrichments and mass flow rates at Russian blending facilities. Prior to its installation in Russian facilities, the BDMS was installed and operated in a UF 6 flow loop in the Paducah Gaseous Diffusion Plant simulating flow and enrichment conditions expected in a typical down-blending facility. A Russian delegation to the US witnessed the equipment demonstration in June, 1998. To conduct the demonstration in the Paducah Gaseous Diffusion Plant (PGDP), the BDMS was required to meet stringent Nuclear Regulatory Commission licensing, safety and operational requirements. The Paducah demonstration was an important milestone in achieving the operational certification for the BDMS use in Russian facilities

Development of a parallel genetic algorithm using MPI and its application in a nuclear reactor core. Design optimization

International Nuclear Information System (INIS)

Waintraub, Marcel; Pereira, Claudio M.N.A.; Baptista, Rafael P.

2005-01-01

This work presents the development of a distributed parallel genetic algorithm applied to a nuclear reactor core design optimization. In the implementation of the parallelism, a 'Message Passing Interface' (MPI) library, standard for parallel computation in distributed memory platforms, has been used. Another important characteristic of MPI is its portability for various architectures. The main objectives of this paper are: validation of the results obtained by the application of this algorithm in a nuclear reactor core optimization problem, through comparisons with previous results presented by Pereira et al.; and performance test of the Brazilian Nuclear Engineering Institute (IEN) cluster in reactors physics optimization problems. The experiments demonstrated that the developed parallel genetic algorithm using the MPI library presented significant gains in the obtained results and an accentuated reduction of the processing time. Such results ratify the use of the parallel genetic algorithms for the solution of nuclear reactor core optimization problems. (author)

Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

International Nuclear Information System (INIS)

Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

2015-01-01

High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

Out-of-core nuclear fuel cycle optimization utilizing an engineering workstation

International Nuclear Information System (INIS)

Turinsky, P.J.; Comes, S.A.

1986-01-01

Within the past several years, rapid advances in computer technology have resulted in substantial increases in their performance. The net effect is that problems that could previously only be executed on mainframe computers can now be executed on micro- and minicomputers. The authors are interested in developing an engineering workstation for nuclear fuel management applications. An engineering workstation is defined as a microcomputer with enhanced graphics and communication capabilities. Current fuel management applications range from using workstations as front-end/back-end processors for mainframe computers to completing fuel management scoping calculations. More recently, interest in using workstations for final in-core design calculations has appeared. The authors have used the VAX 11/750 minicomputer, which is not truly an engineering workstation but has comparable performance, to complete both in-core and out-of-core fuel management scoping studies. In this paper, the authors concentrate on our out-of-core research. While much previous work in this area has dealt with decisions concerned with equilibrium cycles, the current project addresses the more realistic situation of nonequilibrium cycles

77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Science.gov (United States)

2012-05-23

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

International Nuclear Information System (INIS)

Nagano, M.; Sakurai, S.; Yamaguchi, H.

1997-01-01

MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

Beta-ray depth dose in tissue equivalent material due to gaseous radioactive effluents from nuclear power plants

International Nuclear Information System (INIS)

Schadt, W.W.

1978-01-01

The magnitude of the absorbed dose to skin from beta particles emitted by the radionuclides in gaseous effluents from boiling water nuclear power reactors is investigated in this dissertation. Using the radionuclide release patterns of F. Brutschy and the beta dosimetry methods of M. Berger, an equation is derived which gives the dose rate in rads per day when the total radionuclide concentration is one microcurie per gram of air. The coefficients in the equation are presented for a wide range of reactor gas hold-up times (48 minutes to 6 days) and plume environmental transit time (0.5 to 60 minutes). The beta dose rates at the skin surface are found to range from 3.9 to 26.7 rads per day. An upper limit of the relative standard deviation in the dose rate is estimated to be 30 percent. The techniques used to develop the equation are applied to data from the Millstone Nuclear Power Station obtained during the summer of 1972. The beta dose at a site 1.7 miles from the reactor is determined to have been 675 millirads per year at the skin surface and 476 millirads per year at a depth of 200 micrometers. At a site 5.1 miles from the reactor these dose rates were 138 and 100 millirads per year respectively

Nuclear fact book

Energy Technology Data Exchange (ETDEWEB)

Hill, O. F.; Platt, A. M.; Robinson, J. V. [comps

1983-05-01

This reference provides significant highlights and summary facts in the following areas: general energy; nuclear energy; nuclear fuel cycle; uranium supply and enrichment; nuclear reactors; spent fuel and advanced repacking concepts; reprocessing; high-level waste; gaseous waste; transuranic waste; low-level waste; remedial action; transportation; disposal; radiation information; environment; legislation; socio-political aspects; conversion factors; and a glossary. (GHT)

Hybrid modeling approach to improve the forecasting capability for the gaseous radionuclide in a nuclear site

International Nuclear Information System (INIS)

Jeong, Hyojoon; Hwang, Wontae; Kim, Eunhan; Han, Moonhee

2012-01-01

Highlights: ► This study is to improve the reliability of air dispersion modeling. ► Tracer experiments assumed gaseous radionuclides were conducted at a nuclear site. ► The performance of a hybrid modeling combined ISC with ANFIS was investigated.. ► Hybrid modeling approach shows better performance rather than a single ISC model. - Abstract: Predicted air concentrations of radioactive materials are important for an environmental impact assessment for the public health. In this study, the performance of a hybrid modeling combined with the industrial source complex (ISC) model and an adaptive neuro-fuzzy inference system (ANFIS) for predicting tracer concentrations was investigated. Tracer dispersion experiments were performed to produce the field data assuming the accidental release of radioactive material. ANFIS was trained in order that the outputs of the ISC model are similar to the measured data. Judging from the higher correlation coefficients between the measured and the calculated ones, the hybrid modeling approach could be an appropriate technique for an improvement of the modeling capability to predict the air concentrations for radioactive materials.

Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

International Nuclear Information System (INIS)

Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

1990-01-01

It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

International Nuclear Information System (INIS)

Tanabe, Fumiya

2011-01-01

In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

Radioactive clearance discharge of effluent from nuclear and radiation facilities

International Nuclear Information System (INIS)

Liu Xinhua; Xu Chunyan

2013-01-01

On the basis of the basic concepts of radiation safety management system exemption, exclusion and clearance, we expound that the general industrial gaseous and liquid effluent discharges are exempted or excluded, gaseous and liquid effluent discharged from nuclear and radiation facilities are clearance, and non-radioactive. The main purpose of this paper is to clarify the concepts, reach a consensus that the gaseous and liquid effluent discharged from nuclear and radiation facilities are non-radioactive and have no hazard to human health and natural environment. (authors)

Lung Cancer Mortality among Uranium Gaseous Diffusion Plant Workers: A Cohort Study 1952–2004

Directory of Open Access Journals (Sweden)

LW Figgs

2013-07-01

Full Text Available Background: 9%–15% of all lung cancers are attributable to occupational exposures. Reports are disparate regarding elevated lung cancer mortality risk among workers employed at uranium gaseous diffusion plants. Objective: To investigate whether external radiation exposure is associated with lung cancer mortality risk among uranium gaseous diffusion workers. Methods: A cohort of 6820 nuclear industry workers employed from 1952 to 2003 at the Paducah uranium gaseous diffusion plant (PGDP was assembled. A job-specific exposure matrix (JEM was used to determine likely toxic metal exposure categories. In addition, radiation film badge dosimeters were used to monitor cumulative external ionizing radiation exposure. International Classification for Disease (ICD codes 9 and 10 were used to identify 147 lung cancer deaths. Logistic and proportional hazards regression were used to estimate lung cancer mortality risk. Results: Lung cancer mortality risk was elevated among workers who experienced external radiation >3.5 mrem and employment duration >12 years. Conclusion: Employees of uranium gaseous diffusion plants carry a higher risk of lung cancer mortality; the mortality is associated with increased radiation exposure and duration of employment.

Burst shield for a pressurized nuclear-reactor core and method of operating same

International Nuclear Information System (INIS)

Beine, B.; Schilling, F.

1976-01-01

A pressurized nuclear-reactor core stands on a base up from which extends a cylindrical side wall formed of a plurality of hollow iron castings held together by circumferential and longitudinal prestressed elements. A cylindrical space between this shield and the core serves for inspection of the core and is normally filled with cast-iron segmental slabs so that if the core bursts pieces thrown out do not acquire any dangerous kinetic energy before engaging the burst shield. The top of the shield is removably secured to the side so that it can be moved out of the way periodically for removal of the filler slabs and inspection of the core. An anchor on the upper end of each longitudinal prestressing element bears against a sleeve pressing against the uppermost side element, and a nut engageable with this anchor is engageable down over the top to hold it in place, removal of this nut leaving the element prestressed in the side wall. 11 claims, 16 drawing figures

Study on radioactive release of gaseous and liquid effluents during normal operation of AP1000

International Nuclear Information System (INIS)

Gong Quan; Zhou Jing; Liu Yu

2014-01-01

The gaseous and liquid radioactive releases of pressurized water reactors plant during normal operation are an important content of environmental impact assessment and play a significant role in the design of nuclear power plant. According to the design characters of AP1OOO radioactive waste management system and the study on the calculation method and the release pathways, the calculation model of the gaseous and liquid radioactive releases during normal operation for AP1OOO are established. Base on the established calculation model and the design parameters of AP1000, the expected value of gaseous and liquid radioactive releases of AP1OOO is calculated. The results of calculation are compared with the limits in GB 6249-2011 and explain the adder that is included tu account for anticipated operational occurrences, providing a reference for environmental impact assessment of pressurized water reactor. (authors)

Regulatory Audit Activities on Nuclear Design of Reactor Cores

International Nuclear Information System (INIS)

Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

2016-01-01

Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

Regulatory Audit Activities on Nuclear Design of Reactor Cores

Energy Technology Data Exchange (ETDEWEB)

Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

Transfer of gaseous iodine to Tradescantia

International Nuclear Information System (INIS)

Nakamura, Yuji; Ohmomo, Yoichiro.

1984-01-01

Transfer rates of gaseous elemental iodine and methyliodide from atmosphere to Tradescantia were investigated in relation to supposed genetic mutation due to radioactive iodine released from nuclear facilities. The estimated transfer rate of elemental iodine to the young buds of Tradescantia, which was given as the ratio of iodine uptake rate per unit weight of the plant to the concentration of the element in the air, was approximately 7 x 10 -2 cm 3 /g.sec, about 30 to 40 times higher than that of methyliodide. The contribution of direct deposition of elemental iodine was suggested to be significant, although methyliodide was mainly absorbed by respiration through stomata of the plant. (author)

Possible generation of heat from nuclear fusion in Earth's inner core.

Science.gov (United States)

Fukuhara, Mikio

2016-11-23

The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

Core fusion accidents in nuclear power reactors. Knowledge review

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

Energy Technology Data Exchange (ETDEWEB)

Parish, T.A.

1995-03-02

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

A new equation of state Based on Nuclear Statistical Equilibrium for Core-Collapse Simulations

Science.gov (United States)

Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

2012-09-01

We calculate a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores.

Gaseous waste processing facility

International Nuclear Information System (INIS)

Konno, Masanobu; Uchiyama, Yoshio; Suzuki, Kunihiko; Kimura, Masahiro; Kawabe, Ken-ichi.

1992-01-01

Gaseous waste recombiners 'A' and 'B' are connected in series and three-way valves are disposed at the upstream and the downstream of the recombiners A and B, and bypass lines are disposed to the recombiners A and B, respectively. An opening/closing controller for the three-way valves is interlocked with a hydrogen densitometer disposed to a hydrogen injection line. Hydrogen gas and oxygen gas generated by radiolysis in the reactor are extracted from a main condenser and caused to flow into a gaseous waste processing system. Gaseous wastes are introduced together with overheated steams to the recombiner A upon injection of hydrogen. Both of the bypass lines of the recombiners A and B are closed, and recombining reaction for the increased hydrogen gas is processed by the recombiners A and B connected in series. In an operation mode not conducting hydrogen injection, it is passed through the bypass line of the recombiner A and processed by the recombiner B. With such procedures, the increase of gaseous wastes due to hydrogen injection can be coped with existent facilities. (I.N.)

An intelligent nuclear reactor core controller for load following operations, using recurrent neural networks and fuzzy systems

International Nuclear Information System (INIS)

Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.

2003-01-01

In the last decade, the intelligent control community has paid great attention to the topic of intelligent control systems for nuclear plants (core, steam generator...). Papers mostly used approximate and simple mathematical SISO (single-input-single-output) model of nuclear plants for testing and/or tuning of the control systems. They also tried to generalize theses models to a real MIMO (multi-input-multi-output) plant, while nuclear plants are typically of complex nonlinear and multivariable nature with high interactions between their state variables and therefore, many of these proposed intelligent control systems are not appropriate for real cases. In this paper, we designed an on-line intelligent core controller for load following operations, based on a heuristic control algorithm, using a valid and updatable recurrent neural network (RNN). We have used an accurate 3-dimensional core calculation code to represent the real plant and to train the RNN. The results of simulation show that this intelligent controller can control the reactor core during load following operations, using optimum control rod groups manoeuvre and variable overlapping strategy. This methodology represents a simple and reliable procedure for controlling other complex nonlinear MIMO plants, and may improve the responses, comparing to other control systems

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

A genetic algorithm solution for combinatorial problems - the nuclear core reload example

Energy Technology Data Exchange (ETDEWEB)

Schirru, R.; Silva, F.C. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia; Pereira, C.M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Chapot, J.L.C. [FURNAS, Rio de Janeiro, RJ (Brazil)

1997-12-01

This paper presents a solution to Traveling Salesman Problem based upon genetic algorithms (GA), using the classic crossover, but avoiding the feasibility problem in offspring individuals, allowing the natural evolution of the GA without introduction of heuristics in the genetic crossover operator. The genetic model presented, that we call the List Model (LM) is based on the encoding and decoding genotype in the way to always generate a phenotype that has a valid structure, over which will be applied the fitness, represented by the total distance. The main purpose of this work was to develop the basis for a new genetic model to be used in the reload of nuclear core of a PWR. In a generic way, this problem can be interpreted as a a search of the optimal combination of N different fuel elements in N nuclear core `holes`, where each combination or load pattern, determines the neutron flux shape and its associate peak factor. The goal is to find out the load pattern that minimizes the peak factor and consequently maximize the useful life of the nuclear fuel. The GA with the List Model was applied to the Angra-1 PWR reload problem and the results are remarkably better than the ones used in the last fuel cycle. (author). 12 refs., 3 figs., 2 tabs.

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

International Nuclear Information System (INIS)

Samim Anghaie

2002-01-01

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

Studies of gaseous multiplication coefficient in isobutane using a resistive plate chamber

Energy Technology Data Exchange (ETDEWEB)

Goncalves, Josemary A.C.; Botelho, Suzana; Tobias, Carmen C.B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Pontificia Univ. Catolica de Sao Paulo (PUC/SP), SP (Brazil); Lima, Iara B.; Vivaldini, Tulio C. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Ridenti, Marco A.; Pascholati, Paulo R. [Universidade de Sao Paulo (USP), SP (Brazil). Inst. de Fisica. Lab. do Acelerador Linear; Fonte, Paulo; Mangiarotti, Alessio [Universidade de Coimbra (Portugal). Dept. de fisica. Lab. de Instrumentacao e Fisica Experimental de Particulas

2009-07-01

Full text: Due to the increasing demands concerning High Energy Physics, Nuclear Medicine and other Nuclear Applications about gaseous detectors operating in high electric fields, many efforts have been done about the choice of filling gases that fulfill these requirements. In this context, the electron transport parameters in gases, as the gaseous multiplication coefficient, play an important role not only for detector design but also for simulation and modeling of discharges, allowing the validation of electron impact cross-sections. In the present work the preliminary measurements of gaseous multiplication coefficient, as function of the reduced electric field (from 36V/cm.Torr until 93V/cm.Torr), for isobutane are presented. Among several filling gases, isobutane is widely used in resistive plate chambers RPCs, and other gaseous detectors, due to its timing properties. Although its characteristics, there is a lack of swarm parameters data in literature for this gas, mainly at high electric fields. The experimental method used is based on the Pulsed Townsend technique, which follows from Townsend equation solution for a uniform electric field. Considering the ratio between the current (I), measured in avalanche mode, and the primary ionization current (I{sub 0}), the effective multiplication coefficient can be determined, since alpha = d{sup -1}ln(I/I{sub 0}), where d is the gap between the electrodes. In our configuration, the experimental setup consists of two parallel plates enclosure in a stainless steel chamber at gas flow regime. The anode, is made of a high resistivity (2.10{sup 12}{omega}.cm) glass (3mm thick and 14mm diameter), while the cathode is of aluminium (40mm diameter). Primary electrons are produced by irradiating the cathode with a nitrogen laser (LTB MNL200-LD) and are accelerated toward the anode by means of a high voltage power supply (Bertan 225-30). In order to validate the technique and to analyze effects of non-uniformity, results for

Development of conceptual nuclear design of 10MWt research reactor core

International Nuclear Information System (INIS)

Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

2008-03-01

KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

Studies for improvement of regulatory control on the radioactive effluent released from nuclear facilities

International Nuclear Information System (INIS)

Cheong, Jae Hak; Park, H. M.; Song, M. C.; Lee, K. H.; Jang, J. K.; Chun, J. K.; Jeong, K. H.

2005-05-01

This report contains the second-year results of the research project titled 'Studies for Improvement of Regulatory Control on the Radioactive Effluent Released from Nuclear Facilities' and mainly provides technical and strategic approaches to improve performance of regulatory control on the gaseous effluent released from domestic nuclear facilities. The main result contained here includes overview and technical bases of radioactive gaseous effluent control (Chapter 1), reconsideration of the sensitivity requirements for measurement of radioactivity in gaseous effluent sample (Chapter 2), uncertainty analysis of the calculated radioactivity in gaseous effluent (Chapter 3), and improvement of quantification method of noble gas releases (Chapter 4). In addition, analysis of the impact due to combined sampling of particulate from multiple release points (Chapter 5), comparison of domestic nuclear reactors gaseous effluent data to foreign PWRs (Chapter 6), standardized sampling technique for collection of gaseous tritium (Chapter 7), and application of Xe-133 equivalent concept to gaseous effluent control (Chapter 8) are also provided. As a whole, this report provides a generic approach to improve the performance of regulatory control on the gaseous effluent. Therefore, actual enforcement of the recommendations should be preceded by establishment of a series of action plans reflecting on the site- and facility-specific design and operational features

Computation system for nuclear reactor core analysis. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

1977-04-01

This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

Nuclear reactor with several cores

International Nuclear Information System (INIS)

Swars, H.

1977-01-01

Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

Shock absorber in combination with a nuclear reactor core structure

International Nuclear Information System (INIS)

Housman, J.J.

1976-01-01

This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

International Nuclear Information System (INIS)

1977-09-01

After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

Rapid monitoring of gaseous radionuclides using a portable spectrometer

International Nuclear Information System (INIS)

Chung, C.; Tsai, C.H.

1995-01-01

A field gamma ray spectrometer, consisting of a portable high purity germanium detector, portable multichannel analyser, and a notebook computer, was used to conduct in situ rapid scanning of radioactive Ar, Kr and Xe isotopes in the air around a nuclear facility. The portable gamma ray spectrometer was calibrated using Ar, Kr, and Xe radioisotopes, activated in a research reactor and released in a sealed chamber. The unit was further tested inside the reactor containment to monitor the concentration of 41 Ar. In a typical one hour field measurement, the detection limits for some rare gas radionuclides corresponded to dose rates around 0.1 nSv.h -1 , which is far less than the dose rate induced by derived air concentrations imposed by the authority. The dose rate due to ground level concentrations of gaseous radionuclides dispersed from a source, such a nuclear facility or nuclear test, can be monitored in a short period using the portable unit. (Author)

Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

International Nuclear Information System (INIS)

Zamora R, L.; Medina F, A.

1999-01-01

The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

Nuclear power reactor core melt accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

2015-11-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

International Nuclear Information System (INIS)

Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

1993-01-01

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

Energy Technology Data Exchange (ETDEWEB)

Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

2015-07-01

High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

International Nuclear Information System (INIS)

Silva, Alexandro S.; Mazaira, Leorlen Y.R.; Dominguez, Dany S.; Hernandez, Carlos R.G.

2015-01-01

High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Gaseous radioactive waste processing system

International Nuclear Information System (INIS)

Onizawa, Hideo.

1976-01-01

Object: To prevent explosion of hydrogen gas within gaseous radioactive waste by removing the hydrogen gas by means of a hydrogen absorber. Structure: A coolant extracted from a reactor cooling system is sprayed by nozzle into a gaseous phase (hydrogen) portion within a tank, thus causing slipping of radioactive rare gas. The gaseous radioactive waste rich in hydrogen, which is purged in the tank, is forced by a waste gas compressor into a hydrogen occlusion device. The hydrogen occlusion device is filled with hydrogen occluding agents such as Mg, Mg-Ni alloy, V-Nb alloy, La-Ni alloy and so forth, and hydrogen in the waste gas is removed through reaction to produce hydrogen metal. The gaseous radioactive waste, which is deprived of hydrogen and reduced in volume, is stored in an attenuation tank. The hydrogen stored in the hydrogen absorber is released and used again as purge gas. (Horiuchi, T.)

Nuclear many-body problem with repulsive hard core interactions

Energy Technology Data Exchange (ETDEWEB)

Haddad, L M

1965-07-01

The nuclear many-body problem is considered using the perturbation-theoretic approach of Brueckner and collaborators. This approach is outlined with particular attention paid to the graphical representation of the terms in the perturbation expansion. The problem is transformed to centre-of-mass coordinates in configuration space and difficulties involved in ordinary methods of solution of the resulting equation are discussed. A new technique, the 'reference spectrum method', devised by Bethe, Brandow and Petschek in an attempt to simplify the numerical work in presented. The basic equations are derived in this approximation and considering the repulsive hard core part of the interaction only, the effective mass is calculated at high momentum (using the same energy spectrum for both 'particle' and 'hole' states). The result of 0.87m is in agreement with that of Bethe et al. A more complete treatment using the reference spectrum method in introduced and a self-consistent set of equations is established for the reference spectrum parameters again for the case of hard core repulsions. (author)

Development of an automated core model for nuclear reactors

International Nuclear Information System (INIS)

Mosteller, R.D.

1998-01-01

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

Method of removing iodine and compounds thereof from gaseous effluents

International Nuclear Information System (INIS)

Keener, R.L.; Kittle, P.A.

1976-01-01

Anion exchange resins including an acrylic backbone formed by the suspension polymerization of a mixture of an acrylic and a crosslinking monomer are useful in the removal of iodine and iodine compounds from gaseous effluents. Removal of radioactive iodine contaminants, particularly alkyl iodine compounds or hydrogen iodine, under extreme conditions, namely temperatures up to 180 0 C and humidities up to 100 percent, from effluents resulting from a major nuclear accident could probably be adsorbed by these resins described herein

Mass spectrometric study of thermodynamic properties of gaseous lead tellurates. Estimation of formation enthalpies of gaseous lead polonates

Energy Technology Data Exchange (ETDEWEB)

Shugurov, S.M., E-mail: s.shugurov@spbu.ru; Panin, A.I.; Lopatin, S.I.; Emelyanova, K.A.

2016-10-15

Gaseous reactions involving lead oxides, tellurium oxide and lead tellurates were studied by the Knudsen effusion mass spectrometry. Equilibrium constants and reaction enthalpies were evaluated. Structures, molecular parameters and thermodynamic functions of gaseous PbTeO{sub 3} and Pb{sub 2}TeO{sub 4} were calculated by quantum chemistry methods. The formation enthalpies Δ{sub f}H{sup 0} (298.15) = −294 ± 13 for gaseous PbTeO{sub 3} and Δ{sub f}H{sup 0} (298.15) = −499 ± 12 for gaseous Pb{sub 2}TeO{sub 4} were obtained. On the base of these results the formation enthalpies of gaseous PbPoO{sub 3} and Pb{sub 2}PoO{sub 4} were estimated as −249 ± 34 and −478 ± 38, respectively. - Highlights: • Gaseous lead tellurates PbTeO{sub 3}, Pb{sub 2}TeO{sub 4} were discovered. • Their thermodynamic properties were studied using both high temperature mass spectrometry and quantum chemistry computations. • The obtained data allows to predict the formation enthalpies of gaseous lead polonates PbPoO{sub 3}, Pb{sub 2}PoO{sub 4}.

Spent nuclear fuel application of CORE reg-sign systems engineering software

International Nuclear Information System (INIS)

Grimm, R.J.

1996-01-01

The DOE has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF program. The sheer size and complexity of the SNF program has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off the shelf PC based software package, to assist DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

2016-04-15

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

Quantum Effects for a Proton in a Low-Barrier, Double-Well Potential: Core Level Photoemission Spectroscopy of Acetylacetone.

Science.gov (United States)

Feyer, Vitaliy; Prince, Kevin C; Coreno, Marcello; Melandri, Sonia; Maris, Assimo; Evangelisti, Luca; Caminati, Walther; Giuliano, Barbara M; Kjaergaard, Henrik G; Carravetta, Vincenzo

2018-02-01

We have performed core level photoemission spectroscopy of gaseous acetylacetone, its fully deuterated form, and two derivatives, benzoylacetone and dibenzoylmethane. These molecules show intramolecular hydrogen bonds, with a proton located in a double-well potential, whose barrier height is different for the three compounds. This has allowed us to examine the effect of the double-well potential on photoemission spectra. Two distinct O 1s core hole peaks are observed, previously assigned to two chemical states of oxygen. We provide an alternative assignment of the double-peak structure of O 1s spectra by taking full account of the extended nature of the wave function associated with the nuclear motion of the proton, the shape of the ground and final state potentials in which the proton is located, and the nonzero temperature of the samples. The peaks are explained in terms of an unusual Franck-Condon factor distribution.

7th International Symposium on Gaseous Dielectrics

CERN Document Server

James, David

1994-01-01

The Seventh International Symposium on Gaseous Dielectrics was held in Knoxville, Tennessee, U. S. A. , on April 24-28, 1994. The symposium continued the interdisciplinary character and comprehensive approach of the preceding six symposia. Gaseous DielecIries VII is a detailed record of the symposium proceedings. It covers recent advances and developments in a wide range of basic, applied and industrial areas of gaseous dielectrics. It is hoped that Gaseous DielecIries VII will aid future research and development in, and encourage wider industrial use of, gaseous dielectrics. The Organizing Committee of the Seventh International Symposium on Gaseous Dielectrics consisted of G. Addis (U. S. A. ), L. G. Christophorou (U. S. A. ), F. Y. Chu (Canada), A. H. Cookson (U. S. A. ), O. Farish (U. K. ), I. Gallimberti (Italy) , A. Garscadden (U. S. A. ), D. R. James (U. S. A. ), E. Marode (France), T. Nitta (Japan), W. Pfeiffer (Germany), Y. Qiu (China), I. Sauers (U. S. A. ), R. J. Van Brunt (U. S. A. ), and W. Zaengl...

MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

International Nuclear Information System (INIS)

Sugino, K.; Iwai, T.

2006-01-01

MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

International Nuclear Information System (INIS)

Elsawi, Mohamed A.; Alnoamani, Zainab

2015-01-01

The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

The conditions of gaseous fuels development

International Nuclear Information System (INIS)

Anon.

1996-01-01

Face to the actual situation of petrol and gas oil in France, the situation of gaseous fuels appears to be rather modest. However, the aim of gaseous fuels is not to totally supersede the liquid fuels. Such a situation would imply a complete overturn which has not been seriously considered yet. This short paper describes the essential conditions to promote the wider use of gaseous fuels: the intervention of public authorities to adopt a more advantageous tax policy in agreement with the ''Clean Air''law project, a suitable distribution network for gaseous fuels, a choice of vehicles consistent with the urban demand, the development of transformation kits of quality and of dual-fuel vehicles by the car manufacturers. (J.S.)

Position-sensitive gaseous photomultipliers research and applications

CERN Document Server

Francke, Tom; Peskov, Vladimir

2016-01-01

Gaseous photomultipliers are defined as gas-filled devices capable of recording single ultraviolet (UV) and visible photons with high position resolution. Used in a variety of research areas, these detectors can be paired with computers to treat and store imaging information of UV-light. Position-Sensitive Gaseous Photomultipliers: Research and Applications explores the advancement of gaseous detectors as applied for single photon detection. Emphasizing emerging perspectives and new ways to apply gaseous detectors across research fields, this research-based publication is an essential reference source for engineers, physicists, graduate-level students, and researchers.

Ex-vessel molten core debris interactions at CANDU nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

1991-12-31

Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

Ex-vessel molten core debris interactions at CANDU nuclear power plants

International Nuclear Information System (INIS)

Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

1990-01-01

Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDU{sup R} 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

Energy Technology Data Exchange (ETDEWEB)

Mostofian, Sara; Boss, Charles [AECL Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga Ontario L5K 1B2 (Canada)

2008-07-01

In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

Roles of nuclear weak rates on the evolution of degenerate cores in stars

Directory of Open Access Journals (Sweden)

Suzuki Toshio

2017-01-01

Full Text Available Electron-capture and β-decay rates in stellar environments are evaluated with the use of new shell-model Hamiltonians for sd-shell and pf-shell nuclei as well as for nuclei belonging to the island of inversion. Important role of the nuclear weak rates on the final evolution of stellar degenerate cores is presented. The weak interaction rates for sd-shell nuclei are calculated to study nuclear Urca processes in O-Ne-Mg cores of stars with 8-10 M⊙ (solar mass and their effects on the final fate of the stars. Nucleosynthesis of iron-group elements in Type Ia supernova explosions are studied with the weak rates for pf-shell nuclei. The problem of the neutron-rich iron-group isotope over-production compared to the solar abundances is shown to be nearly solved with the use of the new rates and explosion model of slow defraglation with delayed detonation. Evaluation of the weak rates is extended to the island of inversion and the region of neutron-rich nuclei near 78Ni, where two major shells contribute to their configurations.

Development of a multi-objective PBIL evolutionary algorithm applied to a nuclear reactor core reload optimization problem

International Nuclear Information System (INIS)

Machado, Marcelo D.; Dchirru, Roberto

2005-01-01

The nuclear reactor core reload optimization problem consists in finding a pattern of partially burned-up and fresh fuels that optimizes the plant's next operation cycle. This optimization problem has been traditionally solved using an expert's knowledge, but recently artificial intelligence techniques have also been applied successfully. The artificial intelligence optimization techniques generally have a single objective. However, most real-world engineering problems, including nuclear core reload optimization, have more than one objective (multi-objective) and these objectives are usually conflicting. The aim of this work is to develop a tool to solve multi-objective problems based on the Population-Based Incremental Learning (PBIL) algorithm. The new tool is applied to solve the Angra 1 PWR core reload optimization problem with the purpose of creating a Pareto surface, so that a pattern selected from this surface can be applied for the plant's next operation cycle. (author)

An X-ray photoelectron spectroscopy study of the products of the interaction of gaseous IrF6 with fine UO2F2

Directory of Open Access Journals (Sweden)

Prusakov Vladimir N.

2007-01-01

Full Text Available Nuclear fuel reprocessing by fluorination, a dry method of regeneration of spent nuclear fuel, uses UO2F2 for the separation of plutonium from gaseous mixtures. Since plutonium requires special treatment, IrF6 was used as a thermodynamic model of PuF6. The model reaction of the interaction of gaseous IrF6 with fine UO2F2 in the sorption column revealed a change of color of the sorption column contents from pale-yellow to gray and black, indicating the formation of products of such an interaction. The X-ray photoelectron spectroscopy study showed that the interaction of gaseous IrF6 with fine UO2F2 at 125 °C results in the formation of stable iridium compounds where the iridium oxidation state is close to Ir3+. The dependence of the elemental compositions of the layers in the sorption column on the penetration depth of IrF6 was established.

Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

International Nuclear Information System (INIS)

Stroh, K.R.

1979-03-01

The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

Spent nuclear fuel application of CORE reg-sign systems engineering software

International Nuclear Information System (INIS)

Grimm, R.J.

1996-01-01

The Department of Energy (DOE) has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF Program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF Program. The sheer size and complexity of the SNF Program, however, has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off-the-shelf PC based software package, to assist the DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

Restraint system for core elements of a reactor core

International Nuclear Information System (INIS)

Class, G.

1975-01-01

In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

International Nuclear Information System (INIS)

Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

1978-06-01

Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

International Nuclear Information System (INIS)

Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

2000-09-01

A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

The Nuclear Review: the Institution of Nuclear Engineers' response to the Review of Nuclear Power

International Nuclear Information System (INIS)

Anon.

1994-01-01

The United Kingdom Government's Nuclear Review currently underway, addresses whether and in what form nuclear power should continue to be part of the country's power generation capability. This article sets out the response of the Institution of Nuclear Engineers to the Nuclear Review. This pro-nuclear group emphasises the benefits to be gained from diversity of generation in the energy supply industry. The environmentally benign nature of nuclear power is emphasised, in terms of gaseous emissions. The industry's excellent safety record also argues in favour of nuclear power. Finally, as power demand increases globally, a health U.K. nuclear industry could generate British wealth through power exports and via the construction industry. The Institution's view on radioactive waste management is also set out. (UK)

Technical and economic aspects of new gaseous diffusion uranium enrichment capacity

International Nuclear Information System (INIS)

Langley, R.A. Jr.; O'Donnell, A.J.

1977-01-01

Work is well advanced on design and construction of the next major increment of U.S. uranium enrichment capacity. The plant will use the gaseous diffusion process to provide the required capacity and reliability at a competitive enrichment services cost. Gaseous diffusion technology is the base against which other processes are compared in order to assess their commercial viability. While it has generally been described as a mature technology with limited future development potential, work on design of the new U.S. plant has resulted in major improvement in plant design with corresponding decreases in plant capacity and operating costs. The paper describes major technological advances incorporated into the new plant design and their impact on enrichment costs. These include the effects of: - advanced barrier technology; - tandem compressor drive systems; - optimization of number of equipment sizes; - single level plant design; - development of rapid power level change capability; - electrical system simplification; - plant arrangement and layout. Resulting capital costs and projected enrichment costs are summarized. Enrichment costs are placed in the context of total nuclear fuel cycle costs. Trade-offs between uranium feed material quantities and enrichment plant tails assays are described, and optimization of this aspect of the nuclear fuel cycle is discussed. The effect on enrichment plant characteristics is described. Flexibility and capability of the new U.S. enrichment plant to meet these changing optimization conditions are described

Scale model study of the seismic response of a nuclear reactor core

International Nuclear Information System (INIS)

Dove, R.C.; Dunwoody, W.E.; Rhorer, R.L.

1983-01-01

The use of scale models to study the dynamics of a system of graphite core blocks used in certain nuclear reactor designs is described. Scaling laws, material selecton, model instrumentation to measure collision forces, and the response of several models to simulated seismic excitation are covered. The effects of Coulomb friction between the blocks and the clearance gaps between the blocks on the system response to seismic excitation are emphasized

Automated in-core image generation from video to aid visual inspection of nuclear power plant cores

Energy Technology Data Exchange (ETDEWEB)

Murray, Paul, E-mail: paul.murray@strath.ac.uk [Department of Electronic and Electrical Engineering, University of Strathclyde, Technology and Innovation Centre, 99 George Street, Glasgow, G1 1RD (United Kingdom); West, Graeme; Marshall, Stephen; McArthur, Stephen [Dept. Electronic and Electrical Engineering, University of Strathclyde, Royal College Building, 204 George Street, Glasgow G1 1XW (United Kingdom)

2016-04-15

Highlights: • A method is presented which improves visual inspection of reactor cores. • Significant time savings are made to activities on the critical outage path. • New information is extracted from existing data sources without additional overhead. • Examples from industrial case studies across the UK fleet of AGR stations. - Abstract: Inspection and monitoring of key components of nuclear power plant reactors is an essential activity for understanding the current health of the power plant and ensuring that they continue to remain safe to operate. As the power plants age, and the components degrade from their initial start-of-life conditions, the requirement for more and more detailed inspection and monitoring information increases. Deployment of new monitoring and inspection equipment on existing operational plant is complex and expensive, as the effect of introducing new sensing and imaging equipment to the existing operational functions needs to be fully understood. Where existing sources of data can be leveraged, the need for new equipment development and installation can be offset by the development of advanced data processing techniques. This paper introduces a novel technique for creating full 360° panoramic images of the inside surface of fuel channels from in-core inspection footage. Through the development of this technique, a number of technical challenges associated with the constraints of using existing equipment have been addressed. These include: the inability to calibrate the camera specifically for image stitching; dealing with additional data not relevant to the panorama construction; dealing with noisy images; and generalising the approach to work with two different capture devices deployed at seven different Advanced Gas Cooled Reactor nuclear power plants. The resulting data processing system is currently under formal assessment with a view to replacing the existing manual assembly of in-core defect montages. Deployment of the

Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

International Nuclear Information System (INIS)

Mitchell, Jonathan; Fordham, Edmund J.

2014-01-01

Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

Review of coaxial flow gas core nuclear rocket fluid mechanics

International Nuclear Information System (INIS)

Weinstein, H.

1976-01-01

In a prematurely aborted attempt to demonstrate the feasibility of using a gas core nuclear reactor as a rocket engine, NASA initiated a number of studies on the relevant fluid mechanics problems. These studies were carried out at NASA laboratories, universities and industrial research laboratories. Because of the relatively sudden termination of most of this work, a unified overview was never presented which demonstrated the accomplishments of the program and pointed out the areas where additional work was required for a full understanding of the cavity flow. This review attempts to fulfill a part of this need in two important areas

In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

International Nuclear Information System (INIS)

Todt, W.H. Sr.; Playfoot, K.C.

1988-01-01

This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

1985-07-01

The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

International Nuclear Information System (INIS)

Kaya, S.; Yavuz, H.

2001-01-01

A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

Axial power distribution calculation using a neural network in the nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1998-12-31

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

Axial power distribution calculation using a neural network in the nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1997-12-31

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

Interaction between core analysis methodology and nuclear design: some PWR examples

International Nuclear Information System (INIS)

Rothleder, B.M.; Eich, W.J.

1982-01-01

The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

Galvanic cell for processing of used nuclear fuel

Science.gov (United States)

Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

2017-02-07

A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

Electrochemical fluorination for processing of used nuclear fuel

Science.gov (United States)

Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

2016-07-05

A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

Control and balance of nuclear matters used for core fabrication of Super Phenix

International Nuclear Information System (INIS)

Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

1987-05-01

The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

Heat-pipe effect on the transport of gaseous radionuclides released from a nuclear waste container

International Nuclear Information System (INIS)

Zhou, W.; Chambre, P.L.; Pigford, T.H.; Lee, W.W.L.

1990-11-01

When an unsaturated porous medium is subjected to a temperature gradient and the temperature is sufficiently high, vadose water is heated and vaporizes. Vapor flows under its pressure gradient towards colder regions where it condenses. Vaporization and condensation produce a liquid saturation gradient, creating a capillary pressure gradient inside the porous medium. Condensate flows towards the hot end under the influence of a capillary pressure gradient. This is a heat pipe in an unsaturated porous medium. We study analytically the transport of gaseous species released from a spent-fuel waste package, as affected by a time-dependent heat pipe in an unsaturated rock. For parameter values typical of a potential repository in partially saturated fractured tuff at Yucca Mountain, we found that a heat pipe develops shortly after waste is buried, and the heat-pipe's spatial extent is time-dependent. Water vapor movements produced by the heat pipe can significantly affect the migration of gaseous radionuclides. 12 refs., 6 figs., 1 tab

Retention of gaseous isotopes

International Nuclear Information System (INIS)

Yarbro, O.O.; Mailen, J.C.; Stephenson, M.J.

1977-01-01

Retention of gaseous fission products during fuel reprocessing has, in the past, been limited to a modest retention of 131 I when processing fuels decayed less than about 180 days. The projected rapid growth of the nuclear power industry along with a desire to minimize environmental effects is leading to the reassessment of requirements for retention of gaseous fission products, including 131 I, 129 I, 85 Kr, 3 H, and 14 C. Starting in the late 1960s, a significant part of the LMFBR reprocessing development program has been devoted to understanding the behavior of gaseous fission products in plant process and effluent streams and the development of advanced systems for their removal. Systems for iodine control include methods for evolving up to 99% of the iodine from dissolver solutions to minimize its introduction and distribution throughout downstream equipment. An aqueous scrubbing system (Iodox) using 20 M HNO 3 as the scrubbing media effectively removes all significant iodine forms from off-gas streams while handling the kilogram quantities of iodine present in head-end and dissolver off-gas streams. Silver zeolite is very effective for removing iodine forms at low concentration from the larger-volume plant off-gas streams. Removal of iodine from plant liquid effluents by solid sorbents either prior to or following final vaporization appears feasible. Krypton is effectively released during dissolution and can be removed from the relatively small volume head-end and dissolver off-gas stream. Two methods appear applicable for removal and concentration of krypton: (1) selective absorption in fluorocarbons, and (2) cryogenic absorption in liquid nitrogen. The fluorocarbon absorption process appears to be rather tolerant of the normal contaminants (H 2 O, CO 2 , NOsub(x), and organics) present in typical reprocessing plant off-gas whereas the cryogenic system requires an extensive feed gas pretreatment system. Retention of tritium in a reprocessing plant is

Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

International Nuclear Information System (INIS)

Giachetti, R.T.

1989-09-01

This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

Dielectrophoretic separation of gaseous isotopes

International Nuclear Information System (INIS)

McConnell, D.B.

1976-01-01

This invention relates to a process for the separation of gaseous isotopes by electrophoresis assisted by convective countercurrent flow and to an apparatus for use in the process. The invention is especially applicable to heavy water separation from steam; however, it is to be understood that the invention is broadly applicable to the separation of gaseous isotopes having different dipole moments and/or different molecular weights. (author)

A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

International Nuclear Information System (INIS)

Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

2012-01-01

We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ∼ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

Science.gov (United States)

Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

2012-11-01

We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

Energy Technology Data Exchange (ETDEWEB)

Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

2012-11-12

We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

1981-06-01

This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

1981-06-01

This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

Hydrogen and Gaseous Fuel Safety and Toxicity

Energy Technology Data Exchange (ETDEWEB)

Lee C. Cadwallader; J. Sephen Herring

2007-06-01

Non-traditional motor fuels are receiving increased attention and use. This paper examines the safety of three alternative gaseous fuels plus gasoline and the advantages and disadvantages of each. The gaseous fuels are hydrogen, methane (natural gas), and propane. Qualitatively, the overall risks of the four fuels should be close. Gasoline is the most toxic. For small leaks, hydrogen has the highest ignition probability and the gaseous fuels have the highest risk of a burning jet or cloud.

Venturi purification device and its application in purification of gaseous waste of nuclear facilities

International Nuclear Information System (INIS)

Kong Jinsong; Yu Ren; Yang Huanlei

2013-01-01

The working principle of Venturi purification device and its purification of aerosol have been described. Then, taking the gaseous iodine as an example, the absorption process of insoluble gas pollutants is discussed, the calculation methods of the gas-liquid contact area, mass transfer rate and efficiency of mass transfer are educed, and the factors that affect the efficiency of mass transfer are analyzed. (authors)

A study of the annual doses to man from routine gaseous effluent releases of the Philippine Nuclear Power Plant Unit 1 (PNPP-1)

International Nuclear Information System (INIS)

Noriel, M.C.J.

1983-01-01

Individual and population integrated doses from radioactive gaseous releases of the Philippine Nuclear Power Plant 1 (PNPP-1) were calculated using a modified GASPAR Code. Input data consisted of meteorological and site data gathered from the PNPP-1 Final Analysis Report (FASR) population and agricultural data from the National Economic and Development Authority (NEDA) and the National Census and Statistics Office (NCSO). Usage factors were calculated based on Food and Nutrition Research Institute (FNRI) recommended dietary allowances for Filipinos. Results of population integrated dose calculations were used in identifying the critical nuclides, the critical body organs, and the critical pathway. Results from individual dose calculation were used in determining compliance with the dose limits set forth in Appendix D of Part 7 Code of Philippine Atomic Energy Commission (PAEC) regulations. (Author). 23 tabs.; 5 figs

Method of separating tritium contained in gaseous wastes

International Nuclear Information System (INIS)

Hashimoto, Yasuo; Oozono, Hideaki.

1981-01-01

Purpose: To decrease tritium concentration in gaseous wastes to less than the allowable level by removing tritium in gaseous wastes generated upon combustion of radioactive wastes by using a plurality of heat exchangers. Method: Gaseous wastes at high temperature generated upon combustion of radioactive wastes in an incinerator are removed with radioactive solid substances through filters, cooled down to a temperature below 10 0 C by the passage through first and second heat exchangers and decreased with tritium content to less than the allowable concentration by the gaseous wastes at low temperature from the second heat exhcanger. The gaseous wastes at low temperature are used as the cooling medium for the first heat exchanger. They are heat exchanged at the upstream of the second heat exchanger with the cooling water from the third heat exchanger and cooled at the downstream by the cooling water cooled by the cooling medium. The gaseous wastes at low temperature thus cooled below 10 0 C are heated to about 350 0 C in the first heat exchanger and discharged. (Moriyama, K.)

Nuclear physics: the core of matter, the fuel of stars

International Nuclear Information System (INIS)

Schiffer, J.P.

1999-01-01

Dramatic progress has been made in all branches of physics since the National Research Council's 1986 decadal survey of the field. The Physics in a New Era series explores these advances and looks ahead to future goals. The series includes assessments of the major subfields and reports on several smaller subfields, and preparation has begun on an overview volume on the unity of physics, its relationships to other fields, and its contributions to national needs. Nuclear Physics is the latest volume of the series. The book describes current activity in understanding nuclear structure and symmetries, the behavior of matter at extreme densities, the role of nuclear physics in astrophysics and cosmology, and the instrumentation and facilities used by the field. It makes recommendations on the resources needed for experimental and theoretical advances in the coming decade. Nuclear physics addresses the nature of matter making up 99.9 percent of the mass of our everyday world. It explores the nuclear reactions that fuel the stars, including our Sun, which provides the energy for all life on Earth. The field of nuclear physics encompasses some 3,000 experimental and theoretical researchers who work at universities and national laboratories across the United States, as well as the experimental facilities and infrastructure that allow these researchers to address the outstanding scientific questions facing us. This report provides an overview of the frontiers of nuclear physics as we enter the next millennium, with special attention to the state of the science in the United States.The current frontiers of nuclear physics involve fundamental and rapidly evolving issues. One is understanding the structure and behavior of strongly interacting matter in terms of its basic constituents, quarks and gluons, over a wide range of conditions - from normal nuclear matter to the dense cores of neutron stars, and to the Big Bang that was the birth of the universe. Another is to describe

The earths innermost core

International Nuclear Information System (INIS)

Nanda, J.N.

1989-01-01

A new earth model is advanced with a solid innermost core at the centre of the Earth where elements heavier than iron, over and above what can be retained in solution in the iron core, are collected. The innermost core is separated from the solid iron-nickel core by a shell of liquid copper. The innermost core has a natural vibration measured on the earth's surface as the long period 26 seconds microseisms. The earth was formed initially as a liquid sphere with a relatively thin solid crust above the Byerly discontinuity. The trace elements that entered the innermost core amounted to only 0.925 ppm of the molten mass. Gravitational differentiation must have led to the separation of an explosive thickness of pure 235 U causing a fission explosion that could expel beyond the Roche limit a crustal scab which would form the centre piece of the moon. A reservoir of helium floats on the liquid copper. A small proportion of helium-3, a relic of the ancient fission explosion present there will spell the exciting magnetic field. The field is stable for thousands of years because of the presence of large quantity of helium-4 which accounts for most of the gaseous collisions that will not disturb the atomic spin of helium-3 atoms. This field is prone to sudden reversals after long periods of stability. (author). 14 refs

Core of a liquid-cooled nuclear reactor

International Nuclear Information System (INIS)

Wright, J.R.; McFall, A.

1975-01-01

The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

Global physical and numerical stability of a nuclear reactor core

International Nuclear Information System (INIS)

Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

2005-01-01

Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

Non-destructive measurement methods for large scale gaseous diffusion process equipment

International Nuclear Information System (INIS)

Mayer, R.L.; Hagenauer, R.C.; McGinnis, B.R.

1994-01-01

Two measurement methods have been developed to measure non-destructively uranium hold-up in gaseous diffusion plants. These methods include passive neutron and passive γ ray measurements. An additional method, high resolution γ ray spectroscopy, provides supplementary information about additional γ ray emitting isotopes, γ ray correction factors, 235 U/ 234 U ratios and 235 U enrichment. Many of these methods can be used as a general purpose measurement technique for large containers of uranium. Measurement applications for these methods include uranium hold-up, waste measurements, criticality safety and nuclear accountability

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

Science.gov (United States)

Anghaie, S.; Chen, G.

1996-01-01

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

Development of on-line uranium enrichment monitor of gaseous UF6 for uranium enrichment plant

International Nuclear Information System (INIS)

Lu Xuesheng; Liu Guorong; Jin Huimin; Zhao Yonggang; Li Jinghuai; Hao Xueyuan; Ying Bin; Yu Zhaofei

2013-01-01

An on-line enrichment monitor was developed to measure the enrichment of UF 6 , flowing through the processing pipes in uranium enrichment plant. A Nal (Tl) detector was used to measure the count rates of the 185.7 keV γ-ray emitted from 235 U, and the total quantity of uranium was determined from thermodynamic characteristics of gaseous uranium hexafluoride. The results show that the maximum relative standard deviation is less than 1% when the measurement time is 120 s or more and the pressure is more than 2 kPa in the measurement chamber. Uranium enrichment of gaseous uranium hexafluoride in the output end of cascade can be monitored continuously by using the device. It should be effective for nuclear materials accountability verifications and materials balance verification at uranium enrichment plant. (authors)

VISIT - Virtual visits to nuclear power plants

International Nuclear Information System (INIS)

Mollaret, Jean-Christophe

2001-01-01

For more than twenty years, EDFs Communication Division has conducted a policy of opening its generation sites to the general public. Around 300,000 people visit a nuclear power plant every year. However, for the security of persons and the safety of facilities, those parts of the plant situated in controlled areas are not accessible to visitors. For the sake of transparency, EDF has taken an interest in the technologies offered by virtual reality to show the general public what a nuclear power plant is really like, so as to initiate dialogue on nuclear energy, particularly with young people. Visit has been developed with virtual reality technologies. It serves to show the invisible (voyage to the core of fission), the inaccessible and to immerse the visitors in environments which are usually closed to the general public (discovery of the controlled area of a nuclear power plant). Visit is used in Public Information Centres which receive visitors to EDF power plants and during international exhibitions and conferences. Visit allows a virtual tour of the following controlled areas: locker room hot area/cold area, a necessary passage before entering the controlled areas; reactor building; fuel building; waste auxiliary building (liquid, solid and gaseous effluents). It also includes a tour of the rooms or equipment usually accessible to the general public: control room, turbine hall, transformer, air cooling tower

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-01-01

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

Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

Energy Technology Data Exchange (ETDEWEB)

Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

Device for the detection of a gaseous phase inside a nuclear reactor

International Nuclear Information System (INIS)

Marini, Jean; Weilbacher, J.C.

1982-01-01

The detection device includes a multiplicity of ultrasonic emitters secured to the upper side of the closures of at least some of the adaptor tubes so that each emitter receiver directs a beam of ultrasonic sounds inside the vessel in parallel with the axis of the corresponding tube, the latter being fitted with at least one receiver placed inside the tube on the path of the ultrasonic beam. The presence of a gaseous phase in the vessel results in the generation of bubbles some of which reach the inside of the instrumented adaptors. The ultrasonic beam is therefore occulted by this gas, thereby producting a reduced echo [fr

Performance Assessment of Passive Gaseous Provisions (PGAP). Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

2013-07-01

The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000 on the basis of IAEA General Conference resolution GC(44)/RES/21. INPRO helps to ensure the availability of sustainable nuclear energy in the 21st century and seeks to bring together all interested Member States - both technology holders and technology users - to consider joint actions to achieve desired innovations. To contribute to an international consensus on the definition of the reliability of passive systems that involve natural circulation, and on a methodology to assess this reliability, INPRO initiated a collaborative project on Performance Assessment of Passive Gaseous Provisions (PGAP) in 2007. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones, not only to enhance the operational safety of the reactors but also to mitigate the consequences of a severe accident should one occur. However, the reliability of passive safety systems is crucial and must be assessed before they are used extensively in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are a priori unknown. The functions of many passive systems are based on thermohydraulic principles, which until recently were considered as not being subject to any kind of failure. Hence, large and consistent efforts are required to quantify the reliability of such systems. Three participants from three INPRO Member States were involved in this collaborative project. Reliability methods for passive systems (RMPS) and assessment of passive system reliability (APSRA) methodologies were used by the participants to assess the performance and reliability of the passive decay heat removal system of the French gas cooled fast reactor design for station blackout and a loss of coolant accident combined with loss of off-site power, respectively. This publication presents the

Nuclear reactor core

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo; Ishibashi, Yoko; Mochida, Takaaki; Haikawa, Katsumasa; Yamanaka, Akihiro.

1995-01-01

A reactor core is radially divided into an inner region, an outer region and an outermost region. As a fuel, three kinds of fuels, namely, a high enrichment degree fuel at 3.4%, a middle enrichment degree fuel at 2.3% and a low enrichment degree at 1.1% of a fuel average enrichment degree of fission product are used. Each of the fuels is bisected to upper and lower portions at an axial center thereof. The difference of average enrichment degrees between upper and lower portions is 0.1% for the high enrichment degree fuel, 0.3% for the middle enrichment degree fuel and 0.2% for the low enrichment degree fuel. In addition, the composition of fuels in each of radial regions comprises 100% of the low enrichment degree fuels in the outermost region, 91% of the higher enrichment degree fuels and 9% of the middle enrichment degree fuels in the outer region, and 34% of the high enrichment degree fuels and 30% of the middle enrichment degree fuels in the inner region. With such a constitution, fuel economy can be improved while maintaining the thermal margin in an initially loaded reactor core of a BWR type reactor. (I.N.)

Diffusion induced nuclear reactions in metals: a possible source of heat in the core

International Nuclear Information System (INIS)

Hamza, V.M.; Iyer, S.S.S.

1989-01-01

It has recently been proposed that diffusion of light nuclei in metals can give rise to unusual electrical charge distributions in their lattice structures, inducing thereby certain nuclear reactions that are otherwise uncommon. In the light of these results we advance the hypothesis that such nuclear reactions take place in the metal rich core of the earth, based on following observations: 1 - The solubility of hydrogen in metals is relatively high compared to that in silicates. 2 - Studies of rare gas samples in intraplate volcanos and diamonds show that 3 He/ He ratio increases with depth in the mantle. 3 - There are indications that He is positively correlated with enrichment of metals in lavas. We propose that hydrogen incorporated into metallic phases at the time of planetary accretion was carried to the core by downward migration of metal rich melts during the early states of proto-earth. Preliminary estimates suggest that cold fusion reactions can give rise to an average rate of heat generation of 8.2x10 12 W and may thus serve as a supplementary source of energy for the geomagnetic dynamo. (author)

Studies on WWER core diagnostics

International Nuclear Information System (INIS)

Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

1987-01-01

The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

International Nuclear Information System (INIS)

Hecker, H.C.

1984-04-01

This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

3D Core Model for simulation of nuclear power plants: Simulation requirements, model features, and validation

International Nuclear Information System (INIS)

Zerbino, H.

1999-01-01

In 1994-1996, Thomson Training and Simulation (TT and S) earned out the D50 Project, which involved the design and construction of optimized replica simulators for one Dutch and three German Nuclear Power Plants. It was recognized early on that the faithful reproduction of the Siemens reactor control and protection systems would impose extremely stringent demands on the simulation models, particularly the Core physics and the RCS thermohydraulics. The quality of the models, and their thorough validation, were thus essential. The present paper describes the main features of the fully 3D Core model implemented by TT and S, and its extensive validation campaign, which was defined in extremely positive collaboration with the Customer and the Core Data suppliers. (author)

Gaseous phase heat capacity of benzoic acid

NARCIS (Netherlands)

Santos, L.M.N.B.F.; Alves da Rocha, M.A.; Gomes, L.R.; Schröder, B.; Coutinho, J.A.P.

2010-01-01

The gaseous phase heat capacity of benzoic acid (BA) was proven using the experimental technique called the "in vacuum sublimation/vaporization Calvet microcalorimetry drop method". To overcome known experimental shortfalls, the gaseous phase heat capacity of BA monomer was estimated by ab initio

78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Science.gov (United States)

2013-09-12

... 52 [Docket No. PRM-50-105; NRC-2012-0056] In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking; denial...-core thermocouples at different elevations and radial positions throughout the reactor core to enable...

Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

International Nuclear Information System (INIS)

El-Genk, M.S.; Morley, N.J.; Yang, J.Y.

1992-01-01

The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core

Once more about cold nuclear fusion

International Nuclear Information System (INIS)

Brudanin, V.B.; Bystritsky, V.M.; Egorov, V.G.

1989-01-01

The results of the experiments on the search for cold nuclear d-d fusion in chemically pure titanium are given both for electrolysis of heavy water D 2 O and for titanium saturation with gaseous deuterium. The saturation took place at the temperature of 77K and pressure of 50 and 150 atm. A round of experiments with temperature varying from 1 to 600 atm was carried out. The limiting values of the partial rate of the nuclear reaction of d-d fusion with neutron production were obtained per deuteron (at the 95% confidence level): λ f ≤4x10 -25 s -1 (experiment with electrolysis), λ f ≤7x10 -28 s -1 (experiment with gaseous deuterium). 7 refs.; 5 figs.; 2 tabs

Transport properties of gaseous ions over a wide energy range. Part III

International Nuclear Information System (INIS)

Ellis, H.W.; Thackston, M.G.; McDaniel, E.W.; Mason, E.A.

1984-01-01

This paper updates and extends in scope our two previous papers entitled ''Transport Properties of Gaseous Ions over a Wide Energy Range.'' The references to the earlier publications (referred to as ''Part I'' and ''Part II'') are I, H. W. Ellis, R. Y. Pai, E. W. McDonald, E. A. Mason, and L. A. Viehland, ATOMIC DATA AND NUCLEAR DATA TABLES 17, 177--210 (19876); and II, H. W. Ellis, E. W. McDaniel, D. L. Albritton, L. A. Veihland, S. L. Lin, and E. A. Mason, ATOMIC DATA AND NUCLEAR DATA TABLES 22, 179--217 (1978). Parts I and II contained compilations of experimental data on ionic mobilities and diffusion coefficients (both longitudinal and transverse) for ions in neutral gase (almost exclusively at room temperature) in an externally applied electric field

About the application of MCNP4 code in nuclear reactor core design calculations

International Nuclear Information System (INIS)

Svarny, J.

2000-01-01

This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

A nuclear heuristic for application to metaheuristics in-core fuel management optimization

Energy Technology Data Exchange (ETDEWEB)

Meneses, Anderson Alvarenga de Moura, E-mail: ameneses@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program; Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Gambardella, Luca Maria, E-mail: luca@idsia.c [Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Schirru, Roberto, E-mail: schirru@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

2009-07-01

The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

A nuclear heuristic for application to metaheuristics in-core fuel management optimization

International Nuclear Information System (INIS)

Meneses, Anderson Alvarenga de Moura; Gambardella, Luca Maria; Schirru, Roberto

2009-01-01

The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

Assessment of water hammer effects on boiling water nuclear reactor core dynamics

Directory of Open Access Journals (Sweden)

Bousbia-Salah Anis

2007-01-01

Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

Core baffle for nuclear reactors

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1977-01-01

The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

A study on the diffusion of gaseous radioactive effluents based on the statistical method

International Nuclear Information System (INIS)

Na, Man Gyun; Lee, Goung Jin

1998-01-01

A diffusion model of radioactive gaseous effluents is improved to apply for domestic nuclear power plants. Up to now, XOQDOQ computer code package developed by U. S NRC has been used for the assessment of radioactive plume dispersion by normal operation of domestic nuclear power plants. XOQDOQ adopts the straight-line Gaussian plume model which was basically derived for the plane terrain. However, since there are so many mountains in Korea, the several shortcomings of XOQDOQ are improved to consider the complex terrain effects. In this work, wind direction change is considered by modifying the wind rose frequency using meteorological data of the local weather stations. In addition, an effective height correction model, a plume reduction model due to plume penetration into mountain, and a wet deposition model are adopted for more realistic assessments. The proposed methodology is implemented in Yonggwang nuclear power plants, and can be used for other domestic nuclear power plants

3D thermal-hydraulic analysis on core of PWR nuclear power station

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

JSPS-CAS Core University Program seminar on summary of 10-year collaborations in plasma and nuclear fusion research area

International Nuclear Information System (INIS)

Toi, Kazuo; Wang Kongjia

2011-07-01

The JSPS-CAS Core University Program (CUP) seminar on “Summary of 10-year Collaborations in Plasma and Nuclear Fusion Research Area” was held from March 9 to March 11, 2011 in the Okinawa Prefectural Art Museum, Naha city, Okinawa, Japan. The collaboration program on plasma and nuclear fusion started from 2001 under the auspices of Japanese Society of Promotion of Science (JSPS) and Chinese Academy of Sciences (CAS). This year is the last year of the CUP. This seminar was organized in the framework of the CUP. In the seminar, 29 oral talks were presented, having 14 Chinese and 30 Japanese participants. These presentations covered key topics related to the collaboration categories: (1) improvement of core plasma properties, (2) basic research on fusion reactor technologies, and (3) theory and numerical simulation. This seminar aims at summarizing the results obtained through the collaborations for 10 years, and discussing future prospects of China-Japan collaboration in plasma and nuclear fusion research areas. (author)

Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

Energy Technology Data Exchange (ETDEWEB)

Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

1999-07-01

The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

International Nuclear Information System (INIS)

Shirmohammadi, L.; Pazirandeh, A.

2011-01-01

The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

Process for exchanging hydrogen isotopes between gaseous hydrogen and water

International Nuclear Information System (INIS)

Hindin, S. G.; Roberts, G. W.

1980-01-01

A process for exchanging isotopes of hydrogen, particularly tritium, between gaseous hydrogen and water is provided whereby gaseous hydrogen depeleted in tritium and liquid or gaseous water containing tritium are reacted in the presence of a metallic catalyst

Sensitivity of control times in function of core parameters and oscillations control in thermal nuclear systems

International Nuclear Information System (INIS)

Amorim, E.S. do; D'Oliveira, A.B.; Galvao, O.B.; Oyama, K.

1981-03-01

Sensitivity of control times to variation of a thermal reactor core parameters is defined by suitable changes in the power coefficient, core size and fuel enrichment. A control strategy is developed based on control theory concepts and on considerations of the physics of the problem. Digital diffusion theory simulation is described which tends to verify the control concepts considered, face dumped oscillations introduced in one thermal nuclear power system. The effectivity of the control actions, in terms of eliminating oscillations, provided guidelines for the working-group engaged in the analysis of the control rods and its optimal performance. (Author) [pt

A core concept for the self-consistent nuclear energy system based on the promising future technology

International Nuclear Information System (INIS)

Arie, K.; Suzuki, M.; Kawashima, M.; Igashira, M.; Shimizu, A.; Fujii-e, Y.

1995-01-01

Feasibility of FP burning while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated through neutron balance and a fast reactor core. It is shown that all radioactive FPs produced by itself can be burnt by a fast reactor while maintaining breeding capability, assuming separation of radioactive FP and stable FP isotopes. Assuming that the recovery system of fuel and FPs to be burnt is based on a pyro-chemical process, the major long-lived FPs of I, Pd, Tc, Sn, Se can be burnt with keeping breeding capability by suitability arranging materials in the fast reactor core. (Author)

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

International Nuclear Information System (INIS)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Web-based Core Design System Development

International Nuclear Information System (INIS)

Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

2011-01-01

The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

Toward full MOX core design

International Nuclear Information System (INIS)

Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

1999-01-01

This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

Hydrogen production from coal using a nuclear heat source

International Nuclear Information System (INIS)

Quade, R.N.

1977-01-01

A strong candidate for hydrogen production in the intermediate time frame of 1990 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas - a feedstock which may not be available in large quantities in this time frame. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. Bench-scale experimental work on the hydrogasification of coal liquids is being carried out. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed and shows diminishing returns at process temperatures above about 1500 0 F. (author)

Hydrogen production from coal using a nuclear heat source

Science.gov (United States)

Quade, R. N.

1976-01-01

A strong candidate for hydrogen production in the intermediate time frame of 1985 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed, and shows diminishing returns at process temperatures above about 1500 F. A possible scenario combining the relatively abundant and low-cost Western coal deposits with the Gulf Coast hydrogen users is presented which provides high-energy density transportation utilizing coal liquids and uranium.

Methods for reformation of gaseous hydrocarbons using electrical discharge

KAUST Repository

Cha, Min Suk

2017-02-16

Methods for the reformation of gaseous hydrocarbons are provided. The methods can include forming a bubble containing the gaseous hydrocarbon in a liquid. The bubble can be generated to pass in a gap between a pair of electrodes, whereby an electrical discharge is generated in the bubble at the gap between the electrodes. The electrodes can be a metal or metal alloy with a high melting point so they can sustain high voltages of up to about 200 kilovolts. The gaseous hydrocarbon can be combined with an additive gas such as molecular oxygen or carbon dioxide. The reformation of the gaseous hydrocarbon can produce mixtures containing one or more of H2, CO, H2O, CO2, and a lower hydrocarbon such as ethane or ethylene. The reformation of the gaseous hydrocarbon can produce low amounts of CO2 and H2O, e.g. about 15 mol-% or less.

Methods for reformation of gaseous hydrocarbons using electrical discharge

KAUST Repository

Cha, Min; Zhang, Xuming

2017-01-01

Methods for the reformation of gaseous hydrocarbons are provided. The methods can include forming a bubble containing the gaseous hydrocarbon in a liquid. The bubble can be generated to pass in a gap between a pair of electrodes, whereby an electrical discharge is generated in the bubble at the gap between the electrodes. The electrodes can be a metal or metal alloy with a high melting point so they can sustain high voltages of up to about 200 kilovolts. The gaseous hydrocarbon can be combined with an additive gas such as molecular oxygen or carbon dioxide. The reformation of the gaseous hydrocarbon can produce mixtures containing one or more of H2, CO, H2O, CO2, and a lower hydrocarbon such as ethane or ethylene. The reformation of the gaseous hydrocarbon can produce low amounts of CO2 and H2O, e.g. about 15 mol-% or less.

Combination of ozone feed and wet electrostatic precipitator. Experimental study of an innovative system to filter gaseous iodine and iodine containing particles

International Nuclear Information System (INIS)

Gouëllo, Mélany; Hokkinen, Jouni; Kärkelä, Teemu; Auvinen, Ari

2017-01-01

Efficient mitigation systems capable of reducing as much as possible the radioactive discharge to the environment in a severe Nuclear Power Plant (NPP) accident are a necessity. Nuclear Power Plants are equipped with filtration systems which have been characterized for aerosol retention in a short term, but to a far lesser extent for the retention of volatile iodine or for the long-term behaviour of the system. After the Fukushima accident, one of the main concerns of the nuclear industry has been to verify the ability of filtration systems to mitigate the possible source term to the environment. Radiotoxic iodine has a significant contribution to a possible source term in a severe NPP accident. An innovative system has been developed at VTT Technical Research Centre of Finland Ltd. to filter gaseous iodine and iodine containing particles (e.g. I x O y ). The system consists of a combination of an ozone feed and a modern wet electrostatic precipitator (WESP). The electrostatic precipitation (ESP) technique is widely used in the industry to filter out impurities (i.e. particles) in gases. The advantage of a WESP is that the impurities are removed from the system with a solution. They can thus directly be transported to a water container, such as the sump in a nuclear power plant. The addition of ozone feed at the inlet of the WESP allows the oxidation of gaseous iodine into particles. The efficiency of the system for the filtration of gaseous molecular iodine and methyl iodine has been tested and the subsequent observations are further discussed in this paper. The results showed that the combination of WESP together with ozone feed results in a good filtering efficiency against molecular iodine under air atmosphere and air/steam atmosphere. (author)

Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

International Nuclear Information System (INIS)

Sevon, T.

2005-11-01

In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

International Nuclear Information System (INIS)

Eun, Ki Lee.; Yong, Hee Kim.; Jybe, Ho Cha.; Moon, Ghu Park.

2000-01-01

A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

Reactor core of nuclear reactor

International Nuclear Information System (INIS)

Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

1994-01-01

In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

International Nuclear Information System (INIS)

Vukovic, J.; Grgic, D.; Konjarek, D.

2010-01-01

This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

Rift Valley fever phlebovirus NSs protein core domain structure suggests molecular basis for nuclear filaments.

Science.gov (United States)

Barski, Michal; Brennan, Benjamin; Miller, Ona K; Potter, Jane A; Vijayakrishnan, Swetha; Bhella, David; Naismith, James H; Elliott, Richard M; Schwarz-Linek, Ulrich

2017-09-15

Rift Valley fever phlebovirus (RVFV) is a clinically and economically important pathogen increasingly likely to cause widespread epidemics. RVFV virulence depends on the interferon antagonist non-structural protein (NSs), which remains poorly characterized. We identified a stable core domain of RVFV NSs (residues 83-248), and solved its crystal structure, a novel all-helical fold organized into highly ordered fibrils. A hallmark of RVFV pathology is NSs filament formation in infected cell nuclei. Recombinant virus encoding the NSs core domain induced intranuclear filaments, suggesting it contains all essential determinants for nuclear translocation and filament formation. Mutations of key crystal fibril interface residues in viruses encoding full-length NSs completely abrogated intranuclear filament formation in infected cells. We propose the fibrillar arrangement of the NSs core domain in crystals reveals the molecular basis of assembly of this key virulence factor in cell nuclei. Our findings have important implications for fundamental understanding of RVFV virulence.

Decision no 2009-DC-0158 of the 15. of September 2009 by the Nuclear Safety Authority specifying instructions related to modalities of water sampling and consumption, and of release in the environment of liquid and gaseous effluents from the base nuclear installation n. 29 operated by the CIS Bio International on the district of Saclay (Essonne department)

International Nuclear Information System (INIS)

2009-01-01

After references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion), this document specifies the instructions related to modalities of water sampling and consumption, and of release in the environment of liquid and gaseous effluents from the base nuclear installation operated by CIS Bio International in Saclay. These instructions and provisions concern the general means implemented by the operator, records, the control by authorities, the limits of water sampling and consumption, the water sampling process, the sampling works and their maintenance and control, the release of liquid, gaseous and radioactive effluents, their control and their transfer, the limit values for different compounds and species, the control of environment around the site, the information of authorities and of population

Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

Science.gov (United States)

Huddar, Lakshana Ravindranath

With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

Core clamping device for a nuclear reactor

International Nuclear Information System (INIS)

Guenther, R.W.

1974-01-01

The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

Test of active coal capacity for retaining gaseous effluents contaminated by 131 I

International Nuclear Information System (INIS)

Campeanu, Catalina; Cruceru, Madalina; Neacsu, Elena

1999-01-01

The aim of this paper is to test the retaining capacity of the active coal of the filters for gaseous effluents contaminated by 131 I. The active coal filters are design for equipping ventilation installations of nuclear plants in which radioactive iodine and combination of it were produced and handled. Such active coal filters are provided also to sorbonnes for radiochemical and physical handling of radioactive iodine. Testing the retaining capacity of active coal for filters is an obvious step, particularly, when the material was stored long time after preparation

Standardization of radioactive gaseous waste radiation monitoring equipment from the International Electrotechnical Commission

International Nuclear Information System (INIS)

Fitoussi, L.

1983-01-01

After having introduced the works performed by the International Electrotechnical Commission, the authors report on the results of the studies carried out to achieve the standardization of the radiation monitoring equipment concerning the control of the gaseous waste from nuclear installations. The authors present the most interesting aspects of the regulations and recommendations, which include not only a description of the minimum permissible levels of sensitivity for the monitors, but also a description of the set of assays to be established in order to verify the level of performance of the monitors

10 CFR 110.8 - List of nuclear facilities and equipment under NRC export licensing authority.

Science.gov (United States)

2010-01-01

... isotopes of uranium (source material or special nuclear material) including gas centrifuge plants, gaseous diffusion plants, aerodynamic enrichment plants, chemical exchange or ion exchange enrichment plants, laser... of uranium. (See appendices to this part for lists of: gas centrifuge equipment—Appendix B; gaseous...

Process for exchanging hydrogen isotopes between gaseous hydrogen and water

International Nuclear Information System (INIS)

Hindin, S.G.; Roberts, G.W.

1977-01-01

A process is described for exchanging isotopes (particularly tritium) between water and gaseous hydrogen. Isotope depleted gaseous hydrogen and water containing a hydrogen isotope are introduced into the vapour phase in a first reaction area. The steam and gaseous hydrogen are brought into contact with a supported metal catalyst in this area in a parallel flow at a temperature range of around 225 and 300 0 C. An effluent flow comprising a mixture of isotope enriched gaseous hydrogen and depleted steam is evacuated from this area and the steam condensed into liquid water [fr

The Great Deluge Algorithm applied to a nuclear reactor core design optimization problem

International Nuclear Information System (INIS)

Sacco, Wagner F.; Oliveira, Cassiano R.E. de

2005-01-01

The Great Deluge Algorithm (GDA) is a local search algorithm introduced by Dueck. It is an analogy with a flood: the 'water level' rises continuously and the proposed solution must lie above the 'surface' in order to survive. The crucial parameter is the 'rain speed', which controls convergence of the algorithm similarly to Simulated Annealing's annealing schedule. This algorithm is applied to the reactor core design optimization problem, which consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. This problem was previously attacked by the canonical genetic algorithm (GA) and by a Niching Genetic Algorithm (NGA). NGAs were designed to force the genetic algorithm to maintain a heterogeneous population throughout the evolutionary process, avoiding the phenomenon known as genetic drift, where all the individuals converge to a single solution. The results obtained by the Great Deluge Algorithm are compared to those obtained by both algorithms mentioned above. The three algorithms are submitted to the same computational effort and GDA reaches the best results, showing its potential for other applications in the nuclear engineering field as, for instance, the nuclear core reload optimization problem. One of the great advantages of this algorithm over the GA is that it does not require special operators for discrete optimization. (author)

Absorbing method of iodine in radioactive gaseous wastes

International Nuclear Information System (INIS)

Fukutome, Yutaka; Mifuji, Hiroshi; Ito, Sakae.

1983-01-01

Purpose: To maintain an iodine adsorbing efficiency at a high level by keeping the adsorbing atmosphere to more than a predetermined temperature to thereby suppress the degradation and the activity reduction in zeolite. Method: Adsorption and desorption-regeneration of gaseous wastes are performed in parallel by heating gaseous wastes in a heater and switchingly supplying the same to adsorption columns by way of valve operation. Processed gases are cooled in a cooler and desorbed gases are supplied to an after-treatment device to eliminate or recover iodine 131. In the adsorption column, iodine in gaseous wastes is adsorbed to remove by using zeolite, wherein the adsorbing atmosphere is kept at a temperature higher than 40 0 C. This can prevent the formation of an aqueous HNO 3 solution from NO 2 and H 2 O contained in the gaseous wastes and prevent the degradation of the zeolite adsorption layer. (Kawakami, Y.)

Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

2013-10-15

Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

International Nuclear Information System (INIS)

Park, Hae Kyun; Kim, Sang Nyung

2013-01-01

Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

Gaseous Electronics Tables, Atoms, and Molecules

CERN Document Server

Raju, Gorur Govinda

2011-01-01

With the constant emergence of new research and application possibilities, gaseous electronics is more important than ever in disciplines including engineering (electrical, power, mechanical, electronics, and environmental), physics, and electronics. The first resource of its kind, Gaseous Electronics: Tables, Atoms, and Molecules fulfills the author's vision of a stand-alone reference to condense 100 years of research on electron-neutral collision data into one easily searchable volume. It presents most--if not all--of the properly classified experimental results that scientists, researchers,

Order of the 4 March 2014 on the homologation of the decision nr 2014-DC-4016 of the Nuclear Safety Authority of the 16 January 2014 defining the limits of releases in the environment of liquid and gaseous effluents by the basic nuclear installations nr 124, 125, 126 and 137 exploited by EDF-SA in Cattenom (Moselle district)

International Nuclear Information System (INIS)

Durand, J.M.

2014-01-01

This order issued by the French Ministry of ecology, sustainable development and energy is based on a decision taken by the ASN. It concerns the management of pollutions and of the impact of a nuclear installation on the environment: applicable limits of gaseous (notably carbon 14, tritium, rare gases, iodines), chemical, and liquid effluent release. Tables indicate the maximum concentrations

Precision tracking with a single gaseous pixel detector

NARCIS (Netherlands)

Tsigaridas, S.; van Bakel, N.; Bilevych, Y.; Gromov, V.; Hartjes, F.; Hessey, N.P.; de Jong, P.; Kluit, R.

2015-01-01

The importance of micro-pattern gaseous detectors has grown over the past few years after successful usage in a large number of applications in physics experiments and medicine. We develop gaseous pixel detectors using micromegas-based amplification structures on top of CMOS pixel readout chips.

Quantification of cost of margin associated with in-core nuclear fuel management for a PWR

International Nuclear Information System (INIS)

Kropaczek, D.J.; Turinsky, P.J.

1989-01-01

The problem of in-core nuclear fuel management optimization is discussed. The problem is to determine the location of core material, such as the fuel and burnable poisons, so as to minimize (maximize) a stated objective within engineering constraints. Typical objectives include maximization of cycle energy production or discharged fuel exposure, and minimization of power peaking factor or reactor vessel fluence. Constraints include discharge burnup limits and one or more of the possible objectives if not selected as the objective. The optimization problem can be characterized as a large combinatorial problem with nonlinear objective function and constraints, which are likely to be active. The authors have elected to employ the integer Monte Carlo programming method to address this optimization problem because of the just-noted problem characteristics. To evaluate the core physics characteristics as a function of fuel loading pattern, second-order accurate perturbation theory is employed with successive application to improve estimates of the optimum loading pattern. No constraints on fuel movement other than requiring quarter-core symmetry were imposed. In this paper the authors employed this methodology to address a related problem. The problem being addressed can be stated as What is the cost associated with margin? Specifically, they wish to assign some financial value in terms of increased levelized fuel cycle cost associated with an increase in core margin of some type, such as power peaking factor

Decision no 2010-DC-0188 by the Nuclear Safety Authority on the 7. of July 2010 specifying to Electricite de France-Societe Anonyme (EDF-SA) the limits of releases in the environment of liquid and gaseous effluents for the operation of Flamanville 1 (IBN n 108), Flamanville 2 (INB n 109) and Flamanville 3 (INB n 167) reactors

International Nuclear Information System (INIS)

2010-01-01

This document contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents by the three Flamanville nuclear reactors operated by Electricite de France-Societe Anonyme (EDF-SA). Tables present the limits for different gaseous and liquid radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and other beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (boric acid, hydrazine hydrate, morpholine, phosphates, nitrates, various metals, and so on). The limits for thermal releases are also specified

Solar fuel production in a novel polymeric electrolyte membrane photoelectrochemical (PEM-PEC) cell with a web of titania nanotube arrays as photoanode and gaseous reactants

NARCIS (Netherlands)

Stoll, T.; Zafeiropoulos, G.; Tsampas, M. N.

2016-01-01

A novel photoelectrochemical (PEC) cell design is proposed and investigated for H-2 production with gaseous reactants. The core of the cell is a membrane electrode assembly (MEA) that consists of a TiO2 nanotube arrays photoanode, a Pt/C cathode, a Pt/C reference electrode and a proton conducting

The Use of Hidden Markov Models for Anomaly Detection in Nuclear Core Condition Monitoring

Science.gov (United States)

Stephen, Bruce; West, Graeme M.; Galloway, Stuart; McArthur, Stephen D. J.; McDonald, James R.; Towle, Dave

2009-04-01

Unplanned outages can be especially costly for generation companies operating nuclear facilities. Early detection of deviations from expected performance through condition monitoring can allow a more proactive and managed approach to dealing with ageing plant. This paper proposes an anomaly detection framework incorporating the use of the Hidden Markov Model (HMM) to support the analysis of nuclear reactor core condition monitoring data. Fuel Grab Load Trace (FGLT) data gathered within the UK during routine refueling operations has been seen to provide information relating to the condition of the graphite bricks that comprise the core. Although manual analysis of this data is time consuming and requires considerable expertise, this paper demonstrates how techniques such as the HMM can provide analysis support by providing a benchmark model of expected behavior against which future refueling events may be compared. The presence of anomalous behavior in candidate traces is inferred through the underlying statistical foundation of the HMM which gives an observation likelihood averaged along the length of the input sequence. Using this likelihood measure, the engineer can be alerted to anomalous behaviour, indicating data which might require further detailed examination. It is proposed that this data analysis technique is used in conjunction with other intelligent analysis techniques currently employed to analyse FGLT to provide a greater confidence measure in detecting anomalous behaviour from FGLT data.

Entrapment process of radioactive gaseous wastes

International Nuclear Information System (INIS)

Gagneraud, Francis; Gagneraud, Michel.

1981-01-01

Process for collecting chemically inert gaseous radioactive waste in melted substances, whereby the gaseous waste is injected under pressure in a molten substance to its saturation point followed by fast cooling. This substance is constituted of glass, ceramics, metallurgical drosses and slag masses in fusion. Its cooling is carried out by quenching by means of running water or a gas fluid, or by casting into vessels with great thermal inertia such as cast iron or similar, before recovery and confinement in receptacles for storage [fr

Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

International Nuclear Information System (INIS)

Huang Jinhua; Sun Hanhong

1988-08-01

Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

The main methods of solving the problem of radioactive waste management from nuclear power stations and spent fuel reprocessing plants in the USSR

International Nuclear Information System (INIS)

1978-09-01

The main directions of solving the problem of radioactive waste management from nuclear power stations and radiochemical plants, the aspects of gaseous waste management, liquid HLW storage in vessels and the problems of heat removal during storage of vitrified HLW in surface storages are considered. The main problems arising during fine decontamination of gaseous discharges from nuclear power stations and reprocessing plants are discussed. The migration of fission products in the environment and technical aspects of their capture from gaseous discharges are also considered

Feasibility study of passive gamma spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy - low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of gamma-rays from fuel debris

International Nuclear Information System (INIS)

Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

2014-01-01

The technologies applied to the analysis of the Three Mile Island accident were examined in a feasibility study of gamma spectrometry of molten core material from the Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy. The focus is on low-volatile fission products and heavy metal inventory analysis, and the fundamental characteristics of gamma-rays from fuel debris with respect to passive measurements. The inventory ratios of the low-volatile lanthanides, "1"5"4Eu and "1"4"4Ce, to special nuclear materials were evaluated by the entire core inventories in units 1, 2, and 3 with an estimated uncertainty of 9%-13% at the 1σ level for homogenized molten fuel material. The uncertainty is expected to be larger locally owing to the use of the irradiation cycle averaging approach. The ratios were also evaluated as a function of burnup for specific fuel debris with an estimated uncertainty of 13%-25% at the 1σ level for units 1 and 2, and most of the fuels in unit 3, although the uncertainty regarding the separated mixed oxide fuel in unit 3 would be significantly higher owing to the burnup dependence approach. Source photon spectra were also examined and cooling-time-dependent data sets were prepared. The fundamental characteristics of high-energy gamma-rays from fuel debris were investigated by a bare-sphere model transport calculation. Mass attenuation coefficients of fuel debris were evaluated to be insensitive to its possible composition in a high-energy region. The leakage photon ratio was evaluated using a variety of parameters, and a significant impact was confirmed for a certain size of fuel debris. Its correlation was summarized with respect to the leakage photopeak ratio of source "1"5"4Eu. Finally, a preliminary study using a hypothetical canister model of fuel debris based on the experience at Three Mile Island was presented, and future plans were introduced. (author)

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Portsmouth Gaseous Diffusion Plant expansion: final environmental statement. Volume 2. Appendices

International Nuclear Information System (INIS)

1977-09-01

Volume 2 is comprised of appendices: Portsmouth Gaseous Diffusion Plant Existing Facilities; Ecology; Civic Involvement; Social Analysis; Population Projections; Toxicity of Air Pollutants to Biota at Portsmouth Gaseous Diffusion Plant; and Assessment of Noise Effects of an Add-On to the Portsmouth Gaseous Diffusion Plant

Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

International Nuclear Information System (INIS)

Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

2014-01-01

This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

Microbial production of gaseous hydrocarbons

Energy Technology Data Exchange (ETDEWEB)

Fukuda, Hideo

1987-10-20

Microbial production of ethylene, isobutane and a saturated gaseous hydrocarbon mixture was described. Microbial ethylene production was studied with Penicillium digitatum IFO 9372 and a novel pathway of the ethylene biosynthesis through alpha-ketoglutarate was proposed. Rhodotorula minuta IFO 1102 was selected for the microbial production of isobutane and the interesting actions of L-leucine and L-phenylalanine for the isobutane production were found. It was finally presented about the microbial production of a saturated gaseous hydrocarbon mixture with Rhizopus japonicus IFO 4758 was described. A gas mixture was produced through a chemical reaction of SH compounds and some cellular component such as squalene under aerobic conditions. (4 figs, 7 tabs, 41 refs)

Gaseous emissions from coal stockpiles

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-01-15

Stockpiled coal undergoes atmospheric oxidation and desorption processes during open air storage. These processes release gases to the environment which may effect health and safety by their toxicity and flammability. In extreme cases, this could lead to a fire. This report discusses gaseous emissions from coal stockpiles. It covers gas emission mechanisms, and gas sampling and testing methods, before examining in more detail the principal gases that have been emitted. It concludes that there is limited research in this area and more data are needed to evaluate the risks of gaseous emissions. Some methods used to prevent coal self-heating and spontaneous combustion can be applied to reduce emissions from coal stockpiles.

Nuclear energy technology: theory and practice of commercial nuclear power

International Nuclear Information System (INIS)

Knief, R.A.

1982-01-01

Reviews Nuclear Energy Technology: Theory and Practice of Commercial Nuclear Power by Ronald Allen Knief, whose contents include an overview of the basic concepts of reactors and the nuclear fuel cycle; the basics of nuclear physics; reactor theory; heat removal; economics; current concerns at the front and back ends of the fuel cycle; design descriptions of domestic and foreign reactor systems; reactor safety and safeguards; Three Mile Island; and a brief overview of the basic concepts of nuclear fusion. Both magnetic and inertial confinement techniques are clearly outlined. Also reviews Nuclear Fuel Management by Harry W. Graves, Jr., consisting of introductory subjects (e.g. front end of fuel cycle); core physics methodology required for fuel depletion calculations; power capability evaluation (analyzes physical parameters that limit potential core power density); and fuel management topics (economics, loading arrangements and core operation strategies)

Fluid simulation of beryllium transport in the ITER gaseous divertor

International Nuclear Information System (INIS)

Knoll, D.A.; Campbell, R.B.; McHugh, P.R.

1994-01-01

The transport of either intrinsic or injected impurities will play a crucial role in the energy loss mechanisms in the ITER gaseous/cold plasma target divertor. Both 1-D and 2-D multi-charge state fluid codes are used to model the transport of beryllium in the ITER SOL. Our major conclusion is that in order to model the containment of impurities, the background flow field must be known in detail. Comparing 1-D and 2-D solutions, hydrogen flow reversal plays an important role in the entrainment process. Further, the flow of particles from the core plasma also has a strong impact on the resultant entrainment of the impurities in both 1-D and 2-D. It is imperative that those components of poloidal velocity due to E x B and diamagnetic drifts be included in the models. (orig.)

CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

Directory of Open Access Journals (Sweden)

Yohannes Sardjono

2015-04-01

Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

Energy Technology Data Exchange (ETDEWEB)

Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1999-07-01

The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Technical Orders of 10 August 1976 on the limits and procedures applicable to radioactive effluent discharges from nuclear installations

International Nuclear Information System (INIS)

1976-01-01

Seven technical Orders by the competent Ministers (mainly the Ministers of Health, of Industry and Research, of the Quality of Life) lay down the procedures, conditions and limits applicable to gaseous and liquid radioactive effluent discharges from nuclear installations. These Orders of 10 August were published on 12 September 1976 in the Official French Gazette and were made in implementation of the Decree of 6 November 1974 on gaseous radioactive effluent discharges from nuclear installations and the Decree of 31 December 1974 on liquid radioactive effluent discharges from nuclear installations. Apart from the general rules for setting limits and methods for effluent discharges, they specify the measures for environmental monitoring and for control by the Central Service for Protection against Ionizing Radiations. Certain of them contain the general rules for liquid or gaseous effluent discharges from all nuclear installations, while others lay down the rules proper to light water nuclear power plants. Other types of reactor ie. fast breeders are not yet subject to such regulations. (N.E.A.) [fr

Nuclear

International Nuclear Information System (INIS)

2014-01-01

This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

Engineering design of the Aries-IV gaseous divertor

International Nuclear Information System (INIS)

Hasan, M.Z.; Najmabadi, F.; Sharafat, S.

1994-01-01

ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10MPa base pressure. ARIES-IV uses double-null divertors for particle control. Total thermal power recovered from the divertors is 425MW, which is 16% of the total reactor thermal power. Among the desirable goals of divertor design were to avoid the use of tungsten and to use the same structural material and primary coolant as in the blanket design. In order to reduce peak heat flux, the innovative gaseous divertor has been used in ARIES-IV. A gaseous divertor reduces peak heat flux by increasing the surface area and by distributing particle and radiation energy more uniformly. Another benefit of gaseous divertor is the reduction of plasma temperature in the divertor chamber, so that material erosion due to sputtering, can be diminished. This makes the use of low-Z material possible in a gaseous divertor

Growth of graphene films from non-gaseous carbon sources

Science.gov (United States)

Tour, James; Sun, Zhengzong; Yan, Zheng; Ruan, Gedeng; Peng, Zhiwei

2015-08-04

In various embodiments, the present disclosure provides methods of forming graphene films by: (1) depositing a non-gaseous carbon source onto a catalyst surface; (2) exposing the non-gaseous carbon source to at least one gas with a flow rate; and (3) initiating the conversion of the non-gaseous carbon source to the graphene film, where the thickness of the graphene film is controllable by the gas flow rate. Additional embodiments of the present disclosure pertain to graphene films made in accordance with the methods of the present disclosure.

Gaseous isotope correlation technique for safeguards at reprocessing facilities

International Nuclear Information System (INIS)

Ohkubo, Michiaki.

1988-03-01

The isotope correlation technique based on gaseous stable fission products can be used as a means of verifying the input measurement to fuel reprocessing plants. This paper reviews the theoretical background of the gaseous fission product isotope correlation technique. The correlations considered are those between burnup and various isotopic ratios of Kr and Xe nuclides. The feasibility of gaseous ICT application to Pu input accountancy of reprocessing facilities is also discussed. The technique offers the possibility of in situ measurement verification by the inspector. (author). 16 refs, 7 figs

Flux and polarization signals of spatially inhomogeneous gaseous exoplanets

NARCIS (Netherlands)

Karalidi, T.; Stam, D.M.; Guirado, D.

2013-01-01

Aims. We present numerically calculated, disk-integrated, spectropolarimetric signals of starlight that is reflected by vertically and horizontally inhomogeneous gaseous exoplanets. We include various spatial features that are present on Solar System’s gaseous planets: belts and zones, cyclonic

Decision no 2010-DC-0182 by the Nuclear Safety Authority on the 18. of May 2010 specifying the limits of releases in the environment of liquid and gaseous effluents of civil base nuclear installations no 46, no 74 and no 100 operated by Electricite de France-Societe Anonyme (EDF-SA) on the district of Saint-Laurent-Nouan (Loir-et-Cher department)

International Nuclear Information System (INIS)

2010-01-01

This document contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents of civil base nuclear installations operated by Electricite de France-Societe Anonyme (EDF-SA) on the district of Saint-Laurent-Nouan. Commented tables in appendix present the limits for different radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and other alpha, beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (boric acid, hydrazine, morpholine, phosphates, nitrates, various metals, and so on). The limits for thermal releases are also specified

Understanding Core-Collapse Supernovae

Science.gov (United States)

Hix, W. R.; Lentz, E. J.; Baird, M.; Messer, O. E. B.; Mezzacappa, A.; Lee, C.-T.; Bruenn, S. W.; Blondin, J. M.; Marronetti, P.

2010-03-01

Our understanding of core-collapse supernovae continues to improve as better microphysics is included in increasingly realistic neutrino-radiationhydrodynamic simulations. Recent multi-dimensional models with spectral neutrino transport, which slowly develop successful explosions for a range of progenitors between 12 and 25 solar mass, have motivated changes in our understanding of the neutrino reheating mechanism. In a similar fashion, improvements in nuclear physics, most notably explorations of weak interactions on nuclei and the nuclear equation of state, continue to refine our understanding of how supernovae explode. Recent progresses on both the macroscopic and microscopic effects that affect core-collapse supernovae are discussed.

Apparatus for controlling nuclear core debris

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

Disclosed is an apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling

Saclay Center of Nuclear Studies, Direction of Materials and Nuclear Fuels, Department of Physico-Chemistry, Division of Physical Chemistry. 1968 Annual report

International Nuclear Information System (INIS)

Schmidt, M.; Clerc, M.; Le Calve, J.; Bourene, M.; Lesigne, B.; Gillois, M.; Devillers, C.; Arvis, M.; Gilles, L.; Moreau, M.; Sutton, J.; Faraggi, M.; Desalos, J.; Tran Dinh Son; Barat, F.; Hickel, B.; Chachaty, C.; Forchioni, A.; Shiotani, M.; Larher, Y.; Maurice, P.; Le Bail, H.; Nenner, T.

1969-03-01

This document is the 1968 annual report of research activities at the Physico-Chemistry Department (Physical Chemistry Division), part of the Directorate of Materials and Nuclear Fuels of the CEA Saclay center of nuclear studies. The report is divided into two main parts: radiolysis and photolysis studies (gaseous phase, condensed phase), and general physico-chemical studies (sorption, molecular jets)

In core reload design for cycle 4 of Daya Bay nuclear power station both units

International Nuclear Information System (INIS)

Zhang Zongyao; Liu Xudong; Xian Chunyu; Li Dongsheng; Zhang Hong; Liu Changwen; Rui Min; Wang Yingming; Zhao Ke; Zhang Hong; Xiao Min

1998-01-01

The basic principles and the contents of the reload design for Daya Bay nuclear power station are briefly introduced. The in core reload design results, and the comparison between the calculated values and the measured values of both units the fourth cycle are also given. The reload design results of the two units satisfy all the economic requirements and safety criteria. The experimented results shown that the predicated values are tally good with all the measurement values

Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

International Nuclear Information System (INIS)

Frosio, T.; Bonaccorsi, T.; Blaise, P.

2017-01-01

The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

2001-01-01

Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

Fluorination of uranium compounds by gaseous bromine trifluoride and a bromine-fluorine mixture

International Nuclear Information System (INIS)

Sakurai, Tsutomu

1976-03-01

This report summarizes the studies of fluorination of uranium compounds by gaseous BrF 3 and a Br 2 -F 2 mixture, which were carried out in Fluorine Chemistry Laboratory of JAERI in connection with the reprocessing method of nuclear fuels. Although thermodynamically more stable than F 2 , BrF 3 has higher reactivity at relatively low temperatures: fluorination of uranium compounds can be carried out at 100 0 -- 200 0 C by using gaseous BrF 3 . This fluorination temperature is lower than those of F 2 , BrF 5 , ClF and SF 4 , and close to that of ClF 3 . The usage of BrF 3 has however the drawbacks that it requires additional devices to heat the corrosive liquid and to remove Br 2 produced as a byproduct. In order to eliminate the difficulties indicated, a new method of fluorination was developed - the use of a Br 2 -F 2 mixture. Addition of small amounts of Br 2 to the fluorine flow (about 6% in relation to the fluorine concentration) gives marked effects on the rate of fluorination. (auth.)

Gaseous diffusion plant transition from DOE to external regulation

International Nuclear Information System (INIS)

Dann, R.K.; Crites, T.R.; Rahm-Crites, L.K.

1997-01-01

After many years of operation as government-owned/contractor-operated facilities, large portions of the gaseous diffusion plants (GDPs) at Portsmouth, Ohio, and Paducah, Kentucky, were leased to the United States Enrichment Corporation (USEC). These facilities are now certified by the U.S. Nuclear Regulatory Commission (NRC) and subject to oversight by the Occupational Safety and Health Administration (OSHA). The transition from DOE to NRC regulation was more difficult than expected. The original commitment was to achieve NRC certification in October 1995; however, considerably more time was required and transition-related costs escalated. The Oak Ridge Operations Office originally estimated the cost of transition at $60 million; $240 million has been spent to date. The DOE's experience in transitioning the GDPs to USEC operation with NRC oversight provides valuable lessons (both positive and negative) that could be applied to future transitions

Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

International Nuclear Information System (INIS)

Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

1993-03-01

X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

Energy Technology Data Exchange (ETDEWEB)

Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

1993-03-01

X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

Development of advanced nuclear core analysis system applicable to various reactor types (II)

International Nuclear Information System (INIS)

Kaneko, Kunio

2003-03-01

A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

Observation of the state of the nuclear reactor core by means of non-linear observation algorithms

International Nuclear Information System (INIS)

Maciel Palacio, F.E.; Espana, M.D.

1990-01-01

A combined, variable-adaptive structure, non-linear observer was designed in order to observe the state of the nuclear reactor core, based on the Absolute Stability Theory. The observer was proved under noise and modelling error conditions. Successful results were obtained in the observation of the states in both cases, showing clear improvement in the observation due to the application of adaptive and variable structure ideas. (Author) [es

Radioactivity in gaseous waste discharged from the separations facilities during 1978

International Nuclear Information System (INIS)

Anderson, J.D.; Poremba, B.E.

1979-01-01

This document is issued quarterly for the purpose of summarizing the radioactive gaseous wastes that are discharged from the facilities of the Rockwell Hanford Operations. Data on alpha and beta emissions during 1978 are presented where relevant to the gaseous effluent. Emission data are not included on gaseous wastes produced within the 200 Areas by other Hanford contractors

Generation of gaseous tritium standards

International Nuclear Information System (INIS)

Hohorst, F.A.

1994-09-01

The determination of aqueous and non-aqueous tritium in gaseous samples is one type of determination often requested of radioanalytical laboratories. This determination can be made by introducing the sample as a gas into a sampling train containing two silica gel beds separated by.a catalytic oxidizer bed. The first bed traps tritiated water. The sample then passes into and through the oxidizer bed where non-aqueous tritium containing species are oxidized to water and other products of combustion. The second silica gel bed then traps the newly formed tritiated water. Subsequently, silica gel is removed to plastic bottles, deionized water is added, and the mixture is permitted to equilibrate. The tritium content of the equilibrium mixture is then determined by conventional liquid scintillation counting (LSC). For many years, the moisture content of inert, gaseous samples has been determined using monitors which quantitatively electrolyze the moisture present after that moisture has been absorbed by phosphorous pentoxide or other absorbents. The electrochemical reaction is quantitative and definitive, and the energy consumed during electrolysis forms the basis of the continuous display of the moisture present. This report discusses the experimental evaluation of such a monitor as the basis for a technique for conversion of small quantities of SRMs of tritiated water ( 3 HOH) into gaseous tritium standards ( 3 HH)

Decision no 2010-DC-0183 by the Nuclear Safety Authority on the 18. of May 2010 specifying instructions related to modalities of water sampling and consumption, and of release in the environment of liquid and gaseous effluents from the base nuclear installations no 46, no 74 and no 100 operated by Electricite de France-Societe Anonyme (EDF-SA) on the district of Saint-Laurent-Nouan (Loir-et-Cher department)

International Nuclear Information System (INIS)

2010-01-01

After references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion), this document specifies the instructions related to modalities of water sampling and consumption, and of release in the environment of liquid and gaseous effluents from some specific base nuclear installations. These instructions and provisions concern the general means implemented by the operator, records, the control by authorities, the limits of water sampling and consumption, the water sampling process, the sampling works and their maintenance and control, the release of liquid, gaseous and radioactive effluents, their control, and the information of authorities and of population

Decision no 2009-DC-0155 of the 15. of September 2009 by the Nuclear Safety Authority specifying the limits of releases in the environment of liquid and gaseous effluents of base nuclear installations n. 18, 35, 40, 49, 50, 72, 77 and 101 operated by the Atomic Energy Commissariat (CEA) in its Saclay Centre, located on the districts of Saclay, Saint-Aubin and Villiers-le-Bacle (Essonne department)

International Nuclear Information System (INIS)

2009-01-01

This document contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents of base nuclear installations of the Saclay Centre operated by the Atomic Energy Commissariat (CEA). Tables present the limits for different radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and other alpha, beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (HCl, HF, NOx, CO, and so on), as well as limits for thermal releases

Nuclear reactor internals arrangement

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

Energy Technology Data Exchange (ETDEWEB)

Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

2015-07-01

High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

BWR power oscillation evaluation methodologies in core design

International Nuclear Information System (INIS)

Hotta, Akitoshi

1995-01-01

At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

Induced luminescence by charged particles on gaseous, liquid and solid argon

International Nuclear Information System (INIS)

Carvalho Torres, M.J.

1980-01-01

A spectral and a kinetic study of the scintillation induced by β and α particles in gaseous, liquid and solid argon have been made in the wavelength region comprised between 1100 and 3000A. The radiative lifetimes and some spectroscopic parameters of the lowest dimer states ( 1 Σ + sub(u) and 3 Σ + sub(u)) have been determined: tau 0 ( 1 Σ + sub(u)) = 4.2ns; tau 0 ( 3 Σ + sub(u)) = 3.1μs; ΔE( 1 Σ + sub(u)- 3 Σ + sub(u)) = 52 meV; hω = 230 cm -1 . A non radiative de-excitation rate of the 3 Σ + sub(u) state has been measured: approximately 2x10 -17 cm 3 s -1 . By applying an electric field the contribution of the electron-ion recombination mechanism to the gaseous argon scintillation is studied. For condensed argon, the dependence of the ratio between the fluorescence and the phosphorescence intensities on the ionisation power of the impinging particle is verified. The continuum which extends from 1600 to 2900A and that is present only in the gas phase spectra, is ascribed to the radiative de-excitation of molecular ions. A time resolved study of the luminescence of high pressure (1-15atm) argon excited by a pulsed electric discharge has also been performed and is compared with that of the scintillation induced by nuclear particles [fr

Operational experience of gaseous effluent treatment at the Eurochemic reprocessing plant

International Nuclear Information System (INIS)

Osipenco, A.; Detilleux, E.

1977-01-01

The EUROCHEMIC fuel reprocessing plant applies the PUREX flow sheet. Two particular features of the plant influence gaseous and liquid effluents: chemical decanning and the ability to process a wide range of fuels, uranium metal or oxide, having an initial enrichment typical of power reactors (up to 5%) or material testing reactors (up to 93%). The ventilation circuits, treatment plant and monitoring equipment for gaseous releases are briefly described. No retention facilities for rare gases, tritium, or carbon-14 are provided. The releases are monitored for krypton-85, iodine-131, alpha and beta-gamma aerosols and tritium. Between 1966 and 1974 the plant processes about 200 tonnes of power reactor fuel, from which about 0.7 tonnes of plutonium and 1.5 tonnes of highly enriched uranium were separated. The most important points in the operation of the gas cleaning equipment are indicated: efficiency, operational reliability, incidents, etc.. Actual discharges as measured are compared with the limits set in the operation licence. Using the atmospheric diffusion coefficients, the dose commitment is estimated. The low level liquid effluents are passed, after neutralization, to the treatment plant of the Belgian nuclear center CEN/SCK. However, if the activity exceeds the limit set by the CEN/SCK, the effluents are concentrated by evaporation and stored on the EUROCHEMIC site. (orig.) [de

Nuclear reactor core assembly

International Nuclear Information System (INIS)

Baxi, C.B.

1978-01-01

The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

1998-10-01

Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

Minaret, a deterministic neutron transport solver for nuclear core calculations

Energy Technology Data Exchange (ETDEWEB)

Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

2011-07-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Minaret, a deterministic neutron transport solver for nuclear core calculations

International Nuclear Information System (INIS)

Moller, J-Y.; Lautard, J-J.

2011-01-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Reducing the risk to Mars: The gas core nuclear rocket

International Nuclear Information System (INIS)

Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

1998-01-01

The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas-core nuclear rocket (GCNR) has the potential to be such a system. The authors have completed a comparative study of the potential impact that a GCNR could have on a manned Mars mission. The total IMLEO, transit times, and accumulated radiation dose to the crew will be compared with the NASA Design Reference Missions

Fluorine: A key enabling element in the nuclear fuel cycle

OpenAIRE

Crouse, P.L.

2015-01-01

Fluorine - in the form of hydrofluoric acid, anhydrous hydrogen fluoride, elemental gaseous fluorine, fluoropolymers, volatile inorganic fluorides, and more - has played, and still plays, a major role in the nuclear industry. In order to enrich uranium, the metal has to be in the gaseous state. While more exotic methods are known, the standard and most cost-competitive way of achieving this is by means of uranium hexafluoride (UF6). This compound sublimates at low temperatures, and the vapour...

Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

International Nuclear Information System (INIS)

Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

1985-11-01

A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

ROSA full-core and DNBR capabilities

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

2013-01-01

The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

ROSA full-core and DNBR capabilities

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

2012-01-01

This paper presents the latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost two decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

Aerial radiological survey of the area surrounding the Portsmouth Gaseous Diffusion Plant, Portsmouth, Ohio

International Nuclear Information System (INIS)

1979-09-01

An airborne radiological survey was conducted in July 1976. It was centered on the Portsmouth Gaseous Diffusion Plant. Two areas were surveyed: one 35 km 2 and the other 16 km 2 . Using Nal(Tl) scintillation detectors, measurements were made of the terrestrial gamma radiation over the areas with a series of north-south flight lines. The processed data indicated that on-site radioactivity was due to nuclear matterials currently or previously handled, processed, or stored. Off-site activities were found to be due wholly to the naturally occurring 40 K, the 238 U chain, and thorium chain gamma emitters

Attachment of gaseous fission products to aerosols

International Nuclear Information System (INIS)

Skyrme, G.

1985-01-01

Accidents may occur in which the integrity of fuel cladding is breached and volatile fission products are released to the containment atmosphere. In order to assess the magnitude of the subsequent radiological hazard it is necessary to know the transport behaviour of such fission products. It is frequently assumed that the fission products remain in the gaseous phase. There is a possibility, however, that they may attach themselves to particles and hence substantially modify their transport properties. This paper provides a theoretical assessment of the conditions under which gaseous fission products may be attached to aerosol particles. Specific topics discussed are: the mass transfer of a gaseous fission product to an isolated aerosol particle in an infinite medium; the rate at which the concentration of fission products in the gas phase diminishes within a container as a result of deposition on a population of particles; and the distribution of deposited fission product between different particle sizes in a log-normal distribution. It is shown that, for a given mass, small particles are more efficient for fission product attachment, and that only small concentrations of such particles may be necessary to achieve rapid attachment. Conditions under which gaseous fission products are not attached to particles are also considered, viz, the competing processes of deposition onto the containment walls and onto aerosol particles, and the possibility of the removal of aerosols from the containment by various deposition processes, or agglomeration, before attachment takes place. (author)

Initial conceptual design study of self-critical nuclear pumped laser systems

Science.gov (United States)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

The Westinghouse BEACON on-line core monitoring system

International Nuclear Information System (INIS)

Buechel, Robert J.; Boyd, William A.; Casadei, Alberto L.

1995-01-01

BEACON (Best Estimate Analysis of Core Operations - Nuclear), a core monitoring and operational support package developed by Westinghouse, has been installed at many operating PWRs worldwide. The BEACON system is a real-time monitoring system which can be used in plants with both fixed and movable incore detector systems and utilizes an on-line nodal model combined with core instrumentation data to provide continuous core power distribution monitoring. In addition, accurate core-predictive capabilities utilizing a full core nodal model updated according to plant operating history can be made to provide operational support. Core history information is kept and displayed to help operators anticipate core behavior and take pro-active control actions. The BEACON system has been licensed by the U.S. Nuclear Regulatory Commission for direct, continuous monitoring of DNBR and peak linear heat rate. This allows BEACON to be integrated into the plant technical specifications to permit significant relaxation of operating limitations defined by conventional technical specifications. (author). 4 refs, 2 figs, 1 tab

Evaluation of hazards from industrial activities near nuclear power plants

International Nuclear Information System (INIS)

Lannoy, A.; Gobert, T.

1980-01-01

Among the potential hazards which could arise from industrial activity near nuclear power plants, fires and explosions of dangerous products are of particular concern. Indeed, thermal radiation from an adjacent fire could endanger the resistance of a plant's structures. Likewise, an accident explosion would induce an overpressure wave which could affect buildings' integrity. This paper presents the methodology developed by Electricite de France to evaluate the consequences of accidents affecting: - Industrial facilities: refineries, chemical and petrochemical plants, storage areas, pipelines of gaseous, liquid and liquefied materials. - Transportation routes (roads, railways, inland waterways) used to carry dangerous substances (solid explosives, liquid, gaseous or liquefied hydrocarbons). Probabilistic methods have been developed by analysis of actual accident statistics (e.g. risks induced by transportation routes) and realistic and representative accident scenarios have been set up. Five sequences have been identified: Formation of a fluid jet at a breach. Evaporation and possible formation of a liquid layer. Atmospheric dispersion and drift of a gaseous cloud. Heat radiation from fire. Unconfined explosion of a gaseous cloud. This paper gives an overview of the methods and the main assumptions used to deal with each sequence. Those methods, presently applied by Electricite de France, provide a coherent and realistic approach for the evaluation of the risks at nuclear power plants induced by industrial activity. (orig.)

In-core fuel management for the course on operational physics of power reactors

International Nuclear Information System (INIS)

Levine, S.H.

1982-01-01

The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

Nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Prescott, R F

1974-07-11

The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

SMART core preliminary nuclear design-II

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

1997-06-01

Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

Nclear rockets improve the propellant efficiency more than twice compared to CRs and thus significantly reduce the propellant requirement. The superior efficiency of nuclear rockets is due to the combination of the huge energy density and a single low molecular weight propellant utilization. Nuclear Thermal Rockets (NTRs) are particularly suitable for manned missions to Mars because it satisfies a relatively high thrust as well as a high propellant efficiency. NTRs use thermal energy released from a nuclear fission reactor to heat a single low molecular weight propellant, i. e., Hydrogen (H{sub 2}) and then exhausted the extremely heated propellant through a thermodynamic nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (I{sub sp}) which represents the ratio of the thrust over the rate of propellant consumption. The difference of I{sub sp} makes over three times propellant savings of NTRs for a manned Mars mission compared to CRs. NTRs can also be configured to operate bimodally by converting the surplus nuclear energy to auxiliary electric power required for the operation of a spacecraft. Moreover, the concept and technology of NTRs are very simple, already proven, and safe. Thus, NTRs can be applied to various space missions such as solar system exploration, International Space Station (ISS) transport support, Near Earth Objects (NEOs) interception, etc. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The ROK has also begun the research for space nuclear systems as a volunteer of the international space race and a major world nuclear energy country. KANUTER is one of the advanced NTR engines currently under development at KAIST. This bimodal engine is operated in two modes of propulsion with 100 MW

French Regulatory Framework for the Recycling/Reuse of Nuclear Waste and the Dismantling of George Besse Gaseous Diffusion Plant

Energy Technology Data Exchange (ETDEWEB)

Themines, R., E-mail: robert.themines@areva.com [AREVA (France)

2011-07-15

The regulatory framework in France governing the management of materials containing low levels of radionuclides is described. The plans for the management of the materials arising from the dismantling of the Georges Besse Gaseous Diffusion Plant are described as an example of the application of the regulations. (author)

Feasibility study of full-reactor gas core demonstration test

Science.gov (United States)

Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

1973-01-01

Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

Secondary incinerator for radioactive gaseous waste

International Nuclear Information System (INIS)

Takeda, Tadashi; Masuda, Takashi.

1997-01-01

A vessel incorporated with packings, in which at least either of the packings and the vessel is put to induction-heating by high frequency induction coils, is disposed in a flow channel of radioactive gaseous wastes exhausted from a radioactive waste incinerator. The packings include metals such as stainless pipes and electroconductive ceramics such as C-SiC ceramics. Since only electricity is used as an energy source, in the secondary incinerator for the radioactive gaseous wastes, it can be installed in a cell safely. In addition, if ceramics are used, there is no worry of deterioration of the incinerator due to organic materials, and essential functions are not lowered. (T.M.)

Evaluation of BEACON-COLSS Core Monitoring System Benefits

International Nuclear Information System (INIS)

Kim, Joon Sung; Park, Young Ho; Morita, Toshio; Book, Michael A.

2005-01-01

In Korean Standard Nuclear Power Plant COLSS (Core Operating Limit Supervisory System) is used to monitor the DNBR Power Operating Limit (DNBRPOL) and Linear Heat Rate POL (KWPFPOL). Westinghouse and KNFC have developed an upgraded core monitoring system by combining the BEACON TM core monitoring system 1 (Best Estimate Analyzer for Core Operation . Nuclear) and COLSS into an integrated product that is called BEACON-COLSS. BEACON-COLSS generates the 3-D power distribution corrected by the in-core detectors measurements. The 3-D core power distribution methodology in BEACON-COLSS is significantly better than the synthesis methodology in COLSS. BEACONCOLSS uses the CETOP-D 2 thermal hydraulic code instead of CETOP-1. CETOP-D is a multi-channel thermal hydraulics code that will provide more accurate DNBR calculations than the DNBR calculators currently used in COLSS

Gaseous detectors for energy dispersive X-ray fluorescence analysis

Science.gov (United States)

Veloso, J. F. C. A.; Silva, A. L. M.

2018-01-01

The energy resolution capability of gaseous detectors is being used in the last years to perform studies on the detection of characteristic X-ray lines emitted by elements when excited by external radiation sources. One of the most successful techniques is the Energy Dispersive X-ray Fluorescence (EDXRF) analysis. Recent developments in the new generation of micropatterned gaseous detectors (MPGDs), triggered the possibility not only of recording the photon energy, but also of providing position information, extending their application to EDXRF imaging. The relevant features and strategies to be applied in gaseous detectors in order to better fit the requirements for EDXRF imaging will be reviewed and discussed, and some application examples will be presented.

In-core fuel management activities in China

International Nuclear Information System (INIS)

Ruan Keqiang; Chen Renji; Hu Chuanwen

1990-01-01

The development of nuclear power in China has reached such a stage that PWR in-core fuel management becomes an urgent problem. At present the main effort is concentrated on solving the Qinshan nuclear power plant and Daya Bay nuclear power plant fuel management problems. For the Qinshan PWR (300 MWe) two packages of in-core fuel management code were developed, one with simplified nodal diffusion method and the other uses advanced Green's function nodal method. Both were used in the PWR core design. With the help of the two code packages first two cycles of the Qinshan PWR core burn-up were calculated. Besides, several research works are under way in the following areas: improvement of the nodal diffusion method and other coarse mesh method in terms of computing speed and accuracy; backward diffusion technique for fuel management application; optimization technique in the fuel loading pattern searching. As for the Daya Bay PWR plant (twin 900 MWe unit), the problem about using what kind of code package for in-core fuel management is still under discussion. In principle the above mentioned code packages are also applicable to it. Besides PWR, in-core fuel management research works are also under way for research reactors, for example, heavy water research reactor and high flux research reactor in some institutes in China. China also takes active participation in international in-core fuel management activities. (author). 19 refs

Order of the 15. of September 2010 approving the decision no 2010-DC-0188 by the Nuclear Safety Authority on the 7. of July 2010 specifying to Electricite de France-Societe Anonyme (EDF-SA) the limits of releases in the environment of liquid and gaseous effluents for the operation of Flamanville 1 (INB n 108), Flamanville 2 (INB n 109) and Flamanville 3 (INB n 167) reactors