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Sample records for charging machines fission reactor

  1. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  2. Steady-state fusion fission reactor concepts based on stellarator-mirror and mirror machines

    International Nuclear Information System (INIS)

    Neutron sources and hybrid reactors offer a possibility for application of fusion in a not too distant future. Steady-state operation on a time scale of a year without interruption is essential for such applications. In response to this need, our studies are focused on concepts which are not limited by pulsed operation. Special attention is put on mirror machines and a stellarator-mirror concept with localized neutron production. Reactor safety, magnetic coils, power amplification by fission, plasma heating, a radial constant of motion which provides a bounded radial motion in the collision free approximation are some of the issues addressed

  3. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  4. New fission reactor designs

    International Nuclear Information System (INIS)

    A number of critical challenges to the expanded or continued use of nuclear power have developed. These can be categorized as: regulatory restrictions and complications; negative public attitudes; plant complexity; plant life, operations, and maintenance; uncertain load growth, financing; waste management. Solutions to these challenges through advanced reactor design centre around four key technical responses. Passive safety systems are being introduced which use the laws of physics to provide emergency reactor coding, control and shutdown thus eliminating the possibility of human error. Modular construction promises cuts in costs and construction time by shifting the major part of component manufacture from the site to the factory. Standardization also cuts capital costs and in addition operations and repair costs and expedites reactor licensing. Improvements to the fuel cycle include improved fuel types, designs and fabrication, and the reprocessing of and recycling spent fuel back into energy production, thus extending uranium resources and offering a partial solution to the problem of waste disposal. Examples of evolutionary and advanced water-cooled reactors, modular high temperature gas-cooled reactors, and advanced liquid metal cooled fast breeder reactors which are being developed round the world are presented. (author)

  5. Reactor refueling machine simulator

    International Nuclear Information System (INIS)

    This patent describes in combination: a nuclear reactor; a refueling machine having a bridge, trolley and hoist each driven by a separate motor having feedback means for generating a feedback signal indicative of movement thereof. The motors are operable to position the refueling machine over the nuclear reactor for refueling the same. The refueling machine also has a removable control console including means for selectively generating separate motor signals for operating the bridge, trolley and hoist motors and for processing the feedback signals to generate an indication of the positions thereof, separate output leads connecting each of the motor signals to the respective refueling machine motor, and separate input leads for connecting each of the feedback means to the console; and a portable simulator unit comprising: a single simulator motor; a single simulator feedback signal generator connected to the simulator motor for generating a simulator feedback signal in response to operation of the simulator motor; means for selectively connecting the output leads of the console to the simulator unit in place of the refueling machine motors, and for connecting the console input leads to the simulator unit in place of the refueling machine motor feedback means; and means for driving the single simulator motor in response to any of the bridge, trolley or hoist motor signals generated by the console and means for applying the simulator feedback signal to the console input lead associated with the motor signal being generated by the control console

  6. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

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

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

  7. Fission-product burn-up in fast reactors

    International Nuclear Information System (INIS)

    In fast reactors where breeding is emphasized the burn-up of fission products can be of considerable importance. Statistical estimates of fission-product cross-sections are combined with recent yield data for the various fissionable species to estimate the gross fission-product cross-section as a function of irradiation time in a number of fast reactor spectra with various fuels. Because of gaps in yield data for some of the fuel species, it is necessary to interpolate on the yield curves in some cases. The chain yield for a given mass is then apportioned among the chain members through use of the equal charge displacement recipe. The cross-sections estimated for U235 fission products by previous authors are supplemented by estimates for fission products important for other fuels. A range of such spectra is considered. These spectra are characterized by the index (average (Ε-1/2)) in the spectra. The sensitivity of the gross poisoning and its burn-up with respect to spectrum variations are considered. The results are also expressed in terms of a few pseudo-fission products, so that changes in effective cross-section of fission products with irradiation can be taken into account in a simple computational fashion. (author)

  8. Conceptual Analysis of Criticality Aspects of Fission Electric Cell Reactors

    International Nuclear Information System (INIS)

    The U.S. Department of Energy's Nuclear Energy Research Initiative Direct Energy Conversion project has a goal of developing direct energy conversion (DEC) processes suitable for commercial development. DEC is any fission process that returns usable energy with no intermediate thermal process. This project includes the study of the fission electric cell (FEC). In the FEC, fission fragments exit the fuel element cathode and are collected by the cell anode. Previous work [1] has shown the potential of FECs with theoretical efficiencies up to 60%. Inspection of this work indicates the need for additional system criticality studies prior to any conclusions regarding the final FEC reactor configuration. This paper outlines the development of models to facilitate reactor criticality design decisions. The models address criticality, design life, and reactor configuration. In addition, this paper proposes future work to complete the criticality model. The direct energy conversion concept converts nuclear energy to electrical energy without the use of a Carnot cycle based system. Kinetic energy of the highly charged fission fragments is converted directly into electrical potential using strong magnetic fields to separate the positive fission fragments from the electrons that are also produced during the fission process. A parametric analysis is performed using Monte Carlo N-Particle (MCNP) [2] simulations to calculate criticality for exact geometric models. The effect on criticality of changing enrichment, number of cells, size of cells, fuel thickness, and reactor size is determined. Heavy water, helium, and beryllium are each considered for a reflector design. (authors)

  9. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

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

  10. Systems study of tokamak fusion--fission reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-11-01

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

  11. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

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

  12. Nuclear reactor fuelling machine

    International Nuclear Information System (INIS)

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 1800 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 1800 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  13. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  14. Mass and Charge Distribution in Low-Energy Fission

    International Nuclear Information System (INIS)

    The mass and charge distributions for thermal-neutron fission of U235 are discussed in considerable detail and compared with the corresponding distributions in other low-energy fission processes. Points discussed in connection with the mass distributions for binary fission include the positions of the peaks, valley and fine structure in a mass yield curve with respect to filled nuclear shells and the changes in the positions that occur with changing fissioning nucleus and excitation energy. The mass distribution from ternary fission is discussed also. For both binary and ternary fission comments are made concerning the mass distributions of primary fragments (before neutron evaporation) and of fission products (after neutron evaporation). Charge distribution is discussed in terms of charge dispersion among fission products with the same mass number and the variation with mass number of Zp, the ''most probable charge'' (non-integral) for a given mass number. Although direct information about charge distribution is limited to fission products, estimates are presented of charge distribution for primary fission fragments. Knowledge and estimates of mass and charge distribution for a fission process allow estimation of primary yields of all fission products or fragments. Although many estimated primary yields are quite uncertain mainly because of lack of knowledge of charge distribution, especially for fission products formed in low yield; some estimates of primary yields are presented to illustrate the need for and possible practicality of further experimentation. Fission processes other than thermal-neutron fission of U235 that are discussed include thermal-neutron fission of U233 and Pu239, spontaneous fission of Pu240 and Cf252, 14-MeV neutron fission of U235 and U238, 11-MeV proton fission of Ra226 and 22-MeV deuteron fission of Bi209. (author)

  15. Reactor with very low fission product inventory

    International Nuclear Information System (INIS)

    A fast converter with one zone and an internal breeding ratio of 1.00, with liquid fuel in the form of molten plutonium- uranium- and sodium chloride, with a thermal power of 3 GW (th) allows continuous extraction of the volatile fission products (Br, I, Kr, Xe, Te) by means of helium purging in the core. The non-volatile fission products e.g. Sr and Cs can continuously be extracted in a chemical reprocessing plant at the reactor site. The impact on an accidental release of fission products is rather significant; the amounts released are 50-100 times smaller than those in a reference reactor (LWR with oxide fuel). Because the heat sink is relatively large and after heat reduced, the temperature of the fuel does not exceed 5000C after an accident, which greatly reduces the consequences of an accident. (Auth.)

  16. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

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

  17. Fission product decay heat for thermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dickens, J. K.

    1979-01-01

    In the past five years there have been new experimental programs to measure decay heat (i.e., time dependent beta- plus gamma-ray energy release rates from the decay of fission products) following thermal-neutron fission of /sup 235/U, /sup 239/Pu, and /sup 241/Pu for times after fission between 1 and approx. 10/sup 5/ sec. Experimental results from the ORNL program stress the very short times following fission, particularly in the first few hundred sec. Complementing the experimental effort, computer codes have been developed for the computation of decay heat by summation of calculated individual energies released by each one of the fission products. By suitably combining the results of the summation calculations with the recent experimental results, a new Decay Heat Standard has been developed for application to safety analysis of operations of light water reactors. The new standard indicates somewhat smaller energy release rates than those being used at present, and the overall uncertainties assigned to the new standard are much smaller than those being used at present.

  18. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 1013Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 1010Bq (0.5 Ci) per day per ton of fuel

  19. Utilization of fission reactors for fusion engineering testing

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-02-08

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

  20. On the fission chamber pulse charge acquisition and interpretation at MINERVE

    International Nuclear Information System (INIS)

    Fission Chambers (FCs) are widely used as neutron detectors for online flux measurement. The FC current pulse charge is a key observable quantity which depends on specifications such as the filling gas pressure and the FC geometry. In order to study pulse charges, experimental data have been acquired at the Cadarache zero power reactor MINERVE. Two chambers with contrasting specifications have been used. The experimental pulse charge spectrum is interpreted by the mean of a modeling of fission products (FPs) energy deposition within the filling gas.The pulse charge spectrum peaks are found to correspond to FP emitted perpendicularly to the electrodes. (authors)

  1. On the fission chamber pulse charge acquisition and interpretation at MINERVE

    Science.gov (United States)

    Loiseau, P.; Geslot, B.; André, J.

    2013-04-01

    Fission Chambers (FCs) are widely used as neutron detectors for online flux measurement. The FC current pulse charge is a key observable quantity which depends on specifications such as the filling gas pressure and the FC geometry. In order to study pulse charges, experimental data have been acquired at the Cadarache zero power reactor MINERVE. Two chambers with contrasting specifications have been used. The experimental pulse charge spectrum is interpreted by the mean of a modeling of fission products (FPs) energy deposition within the filling gas. The pulse charge spectrum peaks are found to correspond to FP emitted perpendicularly to the electrodes.

  2. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

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

  3. Two-lump fission product model for fast reactor analysis

    International Nuclear Information System (INIS)

    As a part of the Fast-Mixed Spectrum Reactor (FMSR) Project, a study was made on the adequacy of the conventional fission product lump models for the analysis of the different FMSR core concepts. A two-lump fission product model consisting of an odd-A fission product lump and an even-A fission product lump with transmutation between the odd- and even-A lumps was developed. This two-lump model is capable of predicting the exact burnup-dependent behavior of the fission products within a few percent over a wide range of spectra and is therefore also applicable to the conventional fast breeder reactor

  4. Thermal Energetic Reactor with High Reproduction of Fission Materials

    Directory of Open Access Journals (Sweden)

    Vladimir M. Kotov

    2012-01-01

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

  5. Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

    Science.gov (United States)

    Towell, Rusty; Niffte Collaboration

    2015-10-01

    The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

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

    Science.gov (United States)

    Wright, Steven A.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

  8. Multigroup fast fission factor treatment in a thermal reactor lattice

    International Nuclear Information System (INIS)

    A multigroup procedure for the studies of the fast fission effects in the thermal reactor lattice and the calculation of the fast fission factor was developed. The Monte Carlo method and the multigroup procedure were combined to calculate the fast neutron interaction and backscattering effects in a reactor lattice. A set of probabilities calculated by the Monte Carlo method gives a multigroup spectrum of neutrons coming from the moderator and entering the fuel element. Thus, the assumptions adopted so far in defining and calculating the fast fission factor has been avoided, and a new definition including the backscattering and interaction effects in a reactor lattice have been given. (author)

  9. The BR1 Reactor:. a Versatile Tool for Fission Experiments

    Science.gov (United States)

    Wagemans, J.

    2008-04-01

    The BR1 reactor located at the Belgian Nuclear Research Centre SCK·CEN in Mol, Belgium, is a research reactor with a variety of irradiation possibilities. Thanks to its large reactor core, its flexible operation and its different irradiation facilities, this reactor is particularly suited for in-core and ex-core neutron physics experiments. This paper gives a general description of the BR1 reactor, with special emphasis on the available irradiation possibilities. Then some examples of fission experiments that have been performed in the past will be referred to and two ongoing projects related to fission will be presented.

  10. A new method to identify nuclear charges of fission fragments

    International Nuclear Information System (INIS)

    For a mass and velocity selected beam of fission fragments, the elemental components of the beam have been determined by measuring the difference between the time the fragments enter an axial ionization chamber (with the electrical field lines parallel to the particle trajectory) and the time the anode pulse crosses a given level. The nuclear charge resolution achieved for typical fission fragments out of the light mass group in thermal neutron induced fission of 235U is Z/δZ = 43 for a nuclear charge Z = 39. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    D. BELLER; G. POLANSKY; ET AL

    2000-11-01

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

  12. Innovative fission reactors for this century

    International Nuclear Information System (INIS)

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

  13. Nuclear data in the problem of fission reactor decommissioning

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Fission reactor irradiation of candidate ceramics

    International Nuclear Information System (INIS)

    Samples of eleven candidate ceramics (MgO. xAl2O3 in five forms, and Al2O3, Si3N4 and SiC in two forms each) have been sent to the EBR-II fission reactor for neutron irradiation testing. EBR-II is being used because of its high neutron flux and the absence of thermal neutrons which cause undesirable nuclear reactions with 14N. The four families of ceramics selected for inclusion have all shown acceptable performance in irradiation tests carried out by the LASL Ceramics Program or by others. The present study will in some cases yield results at temperatures not yet investigated, and in others will allow the evaluation of forms of these ceramics never before irradiated. Samples will be irradiated at 675 and 825 K to approx. 2.4 x 1022 n/cm2 (E/sub n/ > 0.1 MeV), after which changes in electrical, mechanical, and thermal properties will be determined

  16. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

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

  17. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Development of fission chamber for nuclear reactors controlling

    International Nuclear Information System (INIS)

    Fission Chambers are gas-filled type detectors that operate in the ionization chamber regime, which is without electron multiplication. As the fill-gas is not directly ionized by neutrons, fission chambers are lined with fissile material that through interaction with neutrons fission products are produced, are highly ionizing particles. Pulse type operation of these detectors are used for neutron flux measurements during start up and shut-down reactor conditions in which pulses of high amplitude produced by fission products can be easily discriminated from those produced by alpha radiation from uranium and also from the external gamma field. With current or current fluctuation mode operation (Campbell) the use of these detectors can be extended for the whole range of reactor operation. In this work, it is presented the development and construction of a fission chamber at IPEN-CNEN/SP. Furthermore, the material and techniques used and also the operational characteristics obtained with the first prototype are given. (author)

  20. Refueling machine for a nuclear reactor

    International Nuclear Information System (INIS)

    An improved refuelling machine for inserting and removing fuel assemblies from a nuclear reactor is described which has been designed to increase the reliability of such machines. The system incorporates features which enable the refuelling operation to be performed more efficiently and economically. (U.K.)

  1. Isotopic studies relative to the Oklo natural fission reactors

    International Nuclear Information System (INIS)

    It has been clearly demonstrated that natural fission reactors operated about 2 109 years ago, in rich uranium one deposits of the Oklo mine in the Republique of Gabon. Six reactions zones have been identified in which approximately six tons of 235U were consumed and the same amount of fission products deposited in the ground. These fission products, their filiation isotopes and nuclei formed from neutron captures are precious tracers, which now can be analysed on well localized samples, to obtain informations on the stability in soil of such elements and data on the nuclear parameters and characteristics of the nuclear reactors. The studies which have been developed at Saclay concern several aspects of this phenomenon: the migrations of fission products, the age of the nuclear reaction, the date of the uranium deposit and the temperature of the reaction zones during the operation of the reactors

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

    International Nuclear Information System (INIS)

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

  3. Brief review of the fusion--fission hybrid reactor

    International Nuclear Information System (INIS)

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

  4. distribution of Release Fission Products Through the Nuclear Reactor Site

    International Nuclear Information System (INIS)

    Through the operation of nuclear reactors, radioactive fission products could be release to the environment as a result of severe accidents e.g. Chernobyl accident. Estimation of the atmospheric dispersion, distribution and transport of the radioactive fission products is essential to assessment of the risk to the public from such accidents. In this work, the polluted plume is treated as a matrix of isolated particles.These particles are the fission product isotopes, which compose the radioactive plume.The fission products were classified depending on its half live into three category, long-lived, medium lived and small half-life.The normalized concentrations of the fission product isotopes in the radioactive plume were calculated.The travel time (the time elapsed from the released instant till the deposited time) of each fission products was calculated. The area around the nuclear reactor stack was divided into different zones, started from the reactor stack position until 5 km.The deposited radioactive fission products in each zone was estimated.The calculations were done using the spherical Gaussian plume model

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

    International Nuclear Information System (INIS)

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

  6. Fission yields of molybdenum in the Oklo natural reactor

    International Nuclear Information System (INIS)

    The isotopic compositions of molybdenum in six uranium-rich samples from the Oklo Zone 9 natural reactor were accurately measured by thermal ionization mass spectrometry. The samples were subjected to an ion exchange separation process that removed the isobaric elements zirconium and ruthenium, with high efficiency and a low blank. Molybdenum possesses seven isotopes of which 92,94,96Mo are unaffected by the fission process, enabling the raw data to be corrected for isotope fractionation by normalising to 92Mo/96Mo, and to use 94Mo to correct for the primordial component in each of the fission-produced isotopes. This enables the relative fission yields of Mo to be calculated from the isotopic composition measurements, to give cumulative fission yields of 1:0.941:0.936:1.025 for 95,97,98,100Mo, respectively. These data demonstrate that the most important nuclear process involved in reactor Zone 9 was the thermal neutron fission of 235U. The consistency of the relative cumulative fission yields of all six samples from different locations in the reactor, implies that Mo is a mobile element in the uraninite comprising Zone 9, and that a significant fraction of molybdenum was mobilized within the reactor zone and probably escaped from Zone 9, a conclusion in agreement with earlier published work. (author)

  7. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamic radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  8. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamics radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  9. Experimental studies of fission properties utilized in reactor design

    International Nuclear Information System (INIS)

    Experimental studies of fission properties utilized in reactor design. A programme of experimental studies of fission parameters useful in reactor design is described including the following: (a) The periods and yields of delayed-neutron groups emitted following the neutron-induced fission of Pu241 are measured. Evidence for systematic isotopic dependence of delayed-neutron yields is presented. An experimental investigation of the relation between the time behaviour of delayed-neutron emission and the energy of the incident neutron inducing fission is described. (b) The cross-section for the inducing, of fission in Am243, Pu242 and Pu241 with neutrons in the energy range 0.030 to 1.8 MeV is measured. Emphasis is placed upon the detailed dependence of the fission cross-section on the incident-neutron energy. The absolute values of the cross-sections are given to a precision of ∼25%. (c) Detailed results of a measurement of the Pu241 fission-neutron spectrum are given, including the spectral shape and average fission-neutron energy. Techniques and methods of measuring prompt-fission-neutron spectra are described. (d) The dependence of #-v# (the average number of neutrons emitted per fission) of U235 on the incident neutron energy is measured from 100 keV to 1.6 MeV. #-v# of U238 and other fissile isotopes is compared to #-v# of U235 (thermal). The relative precision of the measurements is #>approx#1.2%. (author)

  10. Langevin description of fission fragment charge distribution from excited nuclei

    CERN Document Server

    Karpov, A V

    2002-01-01

    A stochastic approach to fission dynamics based on a set of three-dimensional Langevin equations was applied to calculate fission-fragment charge distribution of compound nucleus sup 2 sup 3 sup 6 U. The following collective coordinates have been chosen - elongation coordinate, neck-thickness coordinate, and charge-asymmetry coordinate. The friction coefficient of charge mode has been calculated in the framework of one-body and two-body dissipation mechanisms. Analysis of the results has shown that Langevin approach is appropriate for investigation of isobaric distribution. Moreover, the dependences of the variance of the charge distribution on excitation energy and on the two-body viscosity coefficient has been studied

  11. Recent measurements with the NIFFTE fission TPC and the potential to advance thorium fuelled reactors

    International Nuclear Information System (INIS)

    The NIFFTE Fission Time Projection Chamber (TPC) is a powerful tool that is being developed to take precision measurements of neutron-induced fission cross sections of transuranic elements. These improved data are needed for many applications including the development of future generations of nuclear reactors. During the last run at the Los Alamos Neutron Science Center (LANSCE) the fully instrumented TPC took data with several different targets for the first time. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow cross section systematics to be controlled at the level of 1%. Results from this run will be shared. These results are critical to the development of advanced uranium-fuelled reactors. However, there are clear advantages to developing thorium-fuelled reactors including the abundance of thorium verses uranium, minimizing radioactive waste, improved reactor safety, and enhanced proliferation resistance. The potential for using the NIFFTE fission TPC to measure needed cross sections important to the development of thorium fuelled nuclear reactors will also be discussed. (author)

  12. Fission product revaporization in the reactor cooling system

    International Nuclear Information System (INIS)

    The reactor cooling system (RCS) of an LWR can act as an efficient scrubber of volatile fission products released during a meltdown accident before vessel melt-through. This assertion is based on calculations that consider transport of the volatile fission products as vapours or condensed on particles. Retention in the primary system occurs by condensation or reaction with structural surfaces or by fallout of particles containing fission products. It is shown that this picture is perturbed by inclusion of decay heating in the thermal-hydraulic calculations. To do so we make use of the TRAP-MELT3 code which integrates the MERGE and TRAP-MELT2 codes and thus permits simultaneous calculation of thermal-hydraulics and fission product transport in the RCS during the meltdown phase of a severe LWR accident. Calculations on the Surry TMLB' sequence show that while structure temperatures can rise as much as 100 K with inclusion of decay heat, little additional fission product release from the RCS results before melt-through of the reactor vessel. After melt-through, structural temperatures are likely to continue to rise and fission products migrate along the RCS by revolatilizing in the hotter regions and condensing in the cooler regions. The potential for a significant source term of volatile fission products to the containment after melt-through thus exists. For these materials, therefore, the RCS may act more as a retardant than a retainer. Quantification of this conjecture will require further analyses. (author)

  13. Most probable charge of fission products in 24 MeV proton induced fission of 238U

    International Nuclear Information System (INIS)

    The charge distributions of fission products in 24 MeV proton-induced fission of 238U were measured by the use of an ion-guide isotope separator on line. The most probable charge (Zp) of the charge distribution was discussed in view of the charge polarization in the fission process. It was found that Zp mainly lies on the proton-rich side in the light mass region and on the proton-deficient side in the heavy mass region compared with the postulate of the unchanged charge distribution. The charge polarization was examined with respect to production Q values. copyright 1998 The American Physical Society

  14. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

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

  17. Importance of neutron data in fission reactor applications

    International Nuclear Information System (INIS)

    The neutron data required to completely analyze fission reactors includes many isotopes and covers a broad energy range. In both fast and thermal reactors, the neutron inventory is a fine balance determined by the fission properties of 235U, 239Pu and 238U and by the capture cross sections of 238U, fuel materials, structural materials and coolant materials. In fast reactors, the spectrum of neutrons ranges from 1 keV to 3 MeV and is influenced by the elastic and inelastic scattering properties of 238U and the structural and coolant materials. For neutron shielding applications, the important neutron data include the total cross sections of structural and coolant materials in the MeV range. The impact of these basic nuclear data in fission reactor applications is most suitably described by sensitivity analysis. For example, sensitivity coefficients computed for a typical large plutonium fueled fast reactor indicate that a percent increase in the 239Pu(n,f) cross section translates into a 0.59 percent increase in k, a 0.78 percent decrease in the breeding ratio, and 0.71 percent decrease in the sodium coolant reactivity worth. Integral data tests of ENDF/B-IV on the thermal reactor benchmarks indicate that there are no major deficiencies in the H2O and 235U cross sections for thermal systems. However, k/sub eff/ is underpredicted for lattices of slightly enriched systems with an indication that epithermal-to-thermal 238U capture is overpredicted. Fast reactor benchmark tests generally yield a less reactive system for Pu fueled reactors compared to U fueled reactors and the capture rate in 238U relative to the fuel fission rate is generally overpredicted in large systems. Shielding benchmark tests indicate a wide range of deficient neutron data, especially the total elastic and inelastic cross sections of Fe, O, and Na in the high energy range

  18. Fission-reactor experiments for fusion-materials research

    International Nuclear Information System (INIS)

    The US Fusion Materials Program makes extensive use of fission reactors to study the effects of simulated fusion environments on materials and to develop improved alloys for fusion reactor service. The fast reactor, EBR-II, and the mixed spectrum reactors, HFIR and ORR, are all used in the fusion program. The HFIR and ORR produce helium from transmutations of nickel in a two-step thermal neutron absorption reaction beginning with 58Ni, and the fast neutrons in these reactors produce atomic displacements. The simultaneous effects of these phenomena produce damage similar to the very high energy neutrons of a fusion reactor. This paper describes irradiation capsules for mechanical property specimens used in the HFIR and the ORR. A neutron spectral tailoring experiment to achieve the fusion reactor He:dpa ratio will be discussed

  19. A hybrid detector telescope for fission fragments and charged particles

    International Nuclear Information System (INIS)

    Measurement of light charged particle (LCP) multiplicities in coincidence with fission fragments (FFs) during the fusion-fission process is a very useful probe to understand the fission dynamics. In this type of measurement, the LCP's are indented to be measured in a wide range of relative angles (θrel) from 0° to 180° with respect to the FF direction. The conventional method of using two separate detectors one for the FF's and another for the LCPs does not allow to direct the LCPs along the direction of FF (in particular, θrel gas and Egas) and two CsI(Tl)-Si(PIN) detectors mounted at the end of the gas-section. In this paper, the results of in-beam usage of the HDT are presented

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

    International Nuclear Information System (INIS)

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

  1. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

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

  2. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-12-01

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

  3. Most probable charge of fission products in proton-induced fission of 238U and 232Th

    International Nuclear Information System (INIS)

    The charge distributions of fission products in proton-induced fission of 238U and 232Th were measured in a wide mass range. The most probable charges lay on the proton-rich side in the light fragment region and on the proton-deficient side in the heavy one compared with the unchanged charge distribution hypothesis. This result implies that the charge polarization occurs in the fission process. The charge polarization was examined with respect to the ground-state Q values. The estimations by the Q values fairly well reproduced the experimental most probable charges. These results suggest that the fission path to the most favorable charge division may go through the most energetically favorable path at scission point. (author)

  4. Charging machine for the transport of fuel elements

    International Nuclear Information System (INIS)

    Charging machines for the transport of fuel elements for nuclear reactors have got a bridge body supported by two parallel rails via wheels. According to the invention the wheels are fixed to the bridge body by means of guide rods in such a way that at least relative movements in direction of the wheels and transversal to it are possible. Parallel to the guide rods springs and movement attenuators are force-locking by connected. Therefore a stabilizing effect with respect to the transversal forces occurring during earthquakes is achieved. (orig.)

  5. Ceramic materials for fission and fusion nuclear reactors

    International Nuclear Information System (INIS)

    A general survey on the ceramics for nuclear applications is presented. For the fission nuclear reactor, the ceramics materials are almost totally used as fuel e.g. (U,Pu)O2; other types of ceramics, e.g. Uranium-Plutonium carbide and nitride, have been investigated as potential nuclear fuels. The (U,Pu)N compound is to be the fuel for the space nuclear power reactor in the U.S.A. For the fusion nuclear reactor, the ceramics should be the fundamental materials for many components: first wall, breeder, RF heating systems, insulant and shielding parts, etc. In recent years many countries are involved on the research and development of ceramic compounds with the principal purpose of being used in the fusion powerplant (year 2010-2020 ?). An effort has been even made to verify if it is possible to use more ceramic components in the fission nuclear plant (probably differntly disigned) to improve the safety level

  6. Neutrino-driven nucleon fission reactors: Supernovae, quasars, and the big bang

    International Nuclear Information System (INIS)

    The purpose of this work is to establish the existence of naturally occurring celestial neutrino-driven nucleon fission chain reaction reactors as the first step in the development of controlled nucleon fission reactors on Earth. Celestial nucleon fission reactors provide functioning models that serve as starting points for reactor development. Recognizing supernovae, quasars, and the Big Bang as functioning neutrino-driven nucleon fission reactors presents the nuclear industry with a new and significant challenge. That challenge is our technological prowess to achieve a controlled nucleon fission chain reaction using the Earth's resources

  7. Effect of Fission Fragments on the Properties of UO2 Fuel of Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The effect of Xenon (Xe) and (Sr) Strontium fission fragments on the properties of UO2 fuel of pressurized water reactors has been evaluated us ing SRIM-2010 program. The released fission products being highly energetic with different masses, different phase states, and carry different charges cause ionization of the fuel from the surface up to the maximum range with the formation of electron-hole pairs. When the kinetic energy falls below the displacement energy of U and O atoms phonon production takes place. The collision of energetic fission products with the fuel results in the creation of recoil-vacancy pairs. The uranium and oxygen recoils re leased during the collision process changes the oxygen to uranium ratio of the UO2 matrix. The fission fragments as well as the recoils reside in interstitial positions in the structure of UO2 fuel with the result in increasing the internal stresses. The magnitude of damage introduced in the fuel is calculated on the bases of the fission rate of 4% enriched UO2. The released fission fragments and recoils as well as the increase in the fuel temperatures cause swelling of the fuel, increase fuel-clad interaction

  8. Two-billion-year-old nuclear reactors: Nature goes fission

    International Nuclear Information System (INIS)

    Once it was thought that the isotopic composition of natural uranium was invariant. It was thus surprising in 1972 when French scientists observed small but significant deficiencies of the minor isotope 235U in uranium ore. Subsequent investigations traced the isotopically anomalous material to the Oklo mine in the African Republic of Gabon. In the mine, cubic-dekametre-sized pods of rock were found to contain extraordinary concentrations of uranium, as much as 65%, with as little as half the normal isotopic abundance of 235U. In these rocks, neodymium was found to be deficient in the premordial isotope 142Nd and enriched in the fission-produced isotopes 143-150Nd. The presence of fission products was unambiguous evidence that the 235U deficiencies were the result of sustained nuclear fission. Within the heart of the natural reactors, the fission densities were on the order of 1020 fissions/cm3, producing hundreds of megajoules of energy and tens of microwatts of power per gram of rock. Nature had forestalled man's great discovery of energy production by nuclear fission

  9. Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

    International Nuclear Information System (INIS)

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

  10. Time-frequency feature analysis and recognition of fission neutrons signal based on support vector machine

    International Nuclear Information System (INIS)

    Based on the interdependent relationship between fission neutrons (252Cf) and fission chain (235U system), the paper presents the time-frequency feature analysis and recognition in fission neutron signal based on support vector machine (SVM) through the analysis on signal characteristics and the measuring principle of the 252Cf fission neutron signal. The time-frequency characteristics and energy features of the fission neutron signal are extracted by using wavelet decomposition and de-noising wavelet packet decomposition, and then applied to training and classification by means of support vector machine based on statistical learning theory. The results show that, it is effective to obtain features of nuclear signal via wavelet decomposition and de-noising wavelet packet decomposition, and the latter can reflect the internal characteristics of the fission neutron system better. With the training accomplished, the SVM classifier achieves an accuracy rate above 70%, overcoming the lack of training samples, and verifying the effectiveness of the algorithm. (authors)

  11. Applications of nuclear data used in fission reactor monitoring technology

    International Nuclear Information System (INIS)

    Fission reactors continue to play a significant role in the energy, medical, military, analytical, and research activities around the world. To use them effectively, it is necessary to monitor their neutron levels and distributions, their radioactive products, and the reaction rates in their fuel, control, moderator, coolant, structural, and target materials. Nuclear data associated with these materials as well as the materials used as monitors are vitally important in providing data to the operations, experiments, analyses, productions, and surveillance sectors of nuclear technology, science, and engineering. This paper reviews the isotopic abundance, cross-section, decay, and yield data of selected materials and reaction products being applied by the author's laboratory in measurements related to fission reactors

  12. Role of fission gas release in reactor licensing

    International Nuclear Information System (INIS)

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

  13. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Nuclear charge distribution in the spontaneous fission of 252Cf

    CERN Document Server

    Wang, Taofeng; Wang, Liming; Men, Qinghua; Han, Hongyin; Xia, Haihong

    2015-01-01

    The measurement for charge distributions of fragments in 252Cf has been performed by using a unique style of detector setup consisting of a typical grid ionization chamber and a dE-E particle telescope. We found that the fragment mass dependency of the average width of the charge distribution shows a systematic decreased trend with the obvious odd-even effect. The variation of widths of charge distribution with kinetic energies shows an approximate V-shape curve due to the large number of neutron emission for the high excitation energies and cold fragmentation with low excitation energies. As for the behavior of the average nuclear charge with respect to its deviation {\\Delta}Z from the unchanged charge distribution (UCD) as a function of the mass number of primary fragments A*, for asymmetric fission products {\\Delta}Z is negative value, while upon approaching mass symmetry {\\Delta}Z turns positive. Concerning the energy dependence of the most probable charge for given primary mass number A*, the obvious inc...

  16. Nuclear data requirements for fission reactor decommissioning

    International Nuclear Information System (INIS)

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

  17. Fission Yields of Some Isotopes in the Fission of Th232 by Reactor Neutrons

    International Nuclear Information System (INIS)

    The fission yields of the longer-lived isotopes produced in the fission of Th232 are not very well known; existing data show rather large discrepancies and/or uncertainties. Since we feel that at least some of these discrepancies arise from difficulties in measuring the absolute activities of the fission products, we measured the fission yield of 10 selected isotopes whose decay schemes are well understood. The thorium foils were irradiated in a position at the edge of the core of the SAPHIR swimming pool reactor. Following irradiation, the thorium was dissolved after addition of appropriate carriers. The fission products of interest were determined by conventional radiochemical methods that had to be modified slightly to ensure good decontamination from the abundantly formed Pa233 . The chemical yields were determined by gravimetric methods. Counting was done preferentially on a γ-spectrometer that had been calibrated at 11 different energies by standards either obtained from the IAEA or prepared by 4πβ-counting. In the case of Sr90, Ru106 and Ce144 a β-proportional counter was used that had been calibrated for these isotopes. In addition to the sought elements, Mo99 was isolated from each foil to serve as an internal monitor for the number of fissions taking place. The experiment thus gave the ratio of the yield of the sought element to the yield of Mo99. This ratio ''R'' was obtained for Sr90, Ru103, Ru106, Ag111, Pd112, I131, Cs137, Ba140, Ba141, Ce141 and Ce144, Results indicate the existence of a third peak in the yield mass curve in the region of symmetric fission. Yields of fission products relative to the Mo99 yields are given, and the absolute yields calculated by assuming y Mo99 = 2.78%. This number was derived from the work of Iyer et al., and was obtained by normalizing the area under the yield mass curve to 200%. (author)

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

    International Nuclear Information System (INIS)

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

  19. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO2 (1.56 x 10-10 to 7.30 x 10-9 s-1), as well as escape rate constants (7.85 x 10-7 to 3.44 x 10-5 s-1) and diffusion coefficients (3.39 x 10-5 to 4.88 x 10-2 cm2/s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  20. Microscopic modeling of mass and charge distributions in the spontaneous fission of 240Pu

    Science.gov (United States)

    Sadhukhan, Jhilam; Nazarewicz, Witold; Schunck, Nicolas

    2016-01-01

    We propose a methodology to calculate microscopically the mass and charge distributions of spontaneous fission yields. We combine the multidimensional minimization of collective action for fission with stochastic Langevin dynamics to track the relevant fission paths from the ground-state configuration up to scission. The nuclear potential energy and collective inertia governing the tunneling motion are obtained with nuclear density functional theory in the collective space of shape deformations and pairing. We obtain a quantitative agreement with experimental data and find that both the charge and mass distributions in the spontaneous fission of 240Pu are sensitive both to the dissipation in collective motion and to adiabatic fission characteristics.

  1. A Fusion-Fission Reactor Concept based on Viable Technologies

    International Nuclear Information System (INIS)

    Full text: The world needs a great deal of carbon free energy for civilization to continue. Nuclear power is attractive for helping cut carbon emissions and reducing imports of fossil fuel. It is commonly realized that it needs hard work before pure fusion energy could be commercially and economically utilized. Some countries are speeding up the development of their fission industry. In China, the government has decided to develop nuclear power with a mid-term target of ∼40 GWe in 2020. If only PWR is used to meet the huge nuclear capacity requirement, there may be a shortage of fissile uranium and an increase of long-lived nuclear wastes. Therefore, any activity to solve the problems has been welcome. A lot of research activities had been done to evaluate the possibility of the hybrid systems in the world, however, most of them were based on advanced fusion and fission technologies. In this contribution, three types of fusion-fission hybrid reactor concepts, i.e. the energy multiplier named FDS-EM, the fuel breeder named FDS-FB, waste transmuter named FDS-WT, have been proposed for the re-examination of feasibility, capability and safety and environmental potential of fission-fusion hybrid systems. Then based on the re-evaluation activity, a multi-functional fusion-fission reactor concept named FDS-MF simultaneously for nuclear waste transmutation, fissile fuel breeding and thermal energy production based on viable technologies i.e. available or limitedly extrapolated nuclear, processing and fusion technologies is proposed. The tokamak can be designed based on relatively easy-achieved plasma parameters extrapolated from the successful operation of the Experimental Advanced Superconducting Tokamak (EAST) in China and other tokamaks in the world, and the subcritical blanket can be designed based on the well-developed technology of PWR. The design and optimization of fusion plasma core parameters, fission blanket and fuel cycle have been presented. And the

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

    International Nuclear Information System (INIS)

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO2 emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  3. Curved Waveguide Based Nuclear Fission for Small, Lightweight Reactors

    Science.gov (United States)

    Coker, Robert; Putnam, Gabriel

    2012-01-01

    The focus of the presented work is on the creation of a system of grazing incidence, supermirror waveguides for the capture and reuse of fission sourced neutrons. Within research reactors, neutron guides are a well known tool for directing neutrons from the confined and hazardous central core to a more accessible testing or measurement location. Typical neutron guides have rectangular, hollow cross sections, which are crafted as thin, mirrored waveguides plated with metal (commonly nickel). Under glancing angles with incoming neutrons, these waveguides can achieve nearly lossless transport of neutrons to distant instruments. Furthermore, recent developments have created supermirror surfaces which can accommodate neutron grazing angles up to four times as steep as nickel. A completed system will form an enclosing ring or spherical resonator system to a coupled neutron source for the purpose of capturing and reusing free neutrons to sustain and/or accelerate fission. While grazing incidence mirrors are a known method of directing and safely using neutrons, no method has been disclosed for capture and reuse of neutrons or sustainment of fission using a circular waveguide structure. The presented work is in the process of fabricating a functional, highly curved, neutron supermirror using known methods of Ni-Ti layering capable of achieving incident reflection angles up to four times steeper than nickel alone. Parallel work is analytically investigating future geometries, mirror compositions, and sources for enabling sustained fission with applicability to the propulsion and energy goals of NASA and other agencies. Should research into this concept prove feasible, it would lead to development of a high energy density, low mass power source potentially capable of sustaining fission with a fraction of the standard critical mass for a given material and a broadening of feasible materials due to reduced rates of release, absorption, and non-fission for neutrons. This

  4. Nuclear reactor fuel elements charging tool

    International Nuclear Information System (INIS)

    To assist the loading of nuclear reactor fuel elements in a reactor core, positioning blocks with a pyramidal upper face charged to guide the fuel element leg are placed on the lower core plate. A carrier equipped with means of controlled displacement permits movement of the blocks over the lower core plate

  5. Thermohydraulic and nuclear modeling of natural fission reactors

    Science.gov (United States)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SLUTZ,STEPHEN A.; SEIDEL,DAVID B.; POLANSKY,GARY F.; ROCHAU,GARY E.; LIPINSKI,RONALD J.; BESENBRUCH,G.; BROWN,L.C.; PARISH,T.A.; ANGHAIE,S.; BELLER,D.E.

    2000-05-30

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

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

    International Nuclear Information System (INIS)

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

  10. Price of fission product transmutation in power reactors

    International Nuclear Information System (INIS)

    The opportunity of Tc-99 and I-129 transmutation in Russian pressure water VVER-1000 reactor is discussed in this paper. Study of long-lived fission product transmutation shows that if Tc-99 or I-129 are located in VVER-type reactor for a total lifetime, then lifetime and burnup are reduced because of additional capture of neutrons. The reduction is proportional to the incinerated mass of nuclide. Transmutation of either 46.8 kg of Tc-99 or 45.8 kg of I-129 causes a reduction of burnup by 2.25 GW.d/ton that is 5.6 % with respect to the burnup without transmutation. This corresponds to a loss of electric power production of 49 GW.d. If both 42 kg Tc-99 and 43 kg I-129 are transmuted, then reduction of burnup is 4.51 GW.d/ton that is 11.3 % of the burnup without transmutation. Loss of electric power production is 99 GW.d. The result does not practically depend on way of transmuted nuclide placement. This loss of power is a price that should be paid for transmutation of nuclides without their removal during reactor operation. One would avoid the reduction of burnup and power loss if it would be possible to find such way of nuclide placement for irradiation in reactor, which would permit to extract nuclides from operating reactor a certain time before next fuel reloading

  11. A fission fragment reactor concept for nuclear space propulsion

    Science.gov (United States)

    Suo-Anttila, A. J.; Parma, E. J.; Wright, S. A.; Vernon, M. E.; Pickard, P. S.

    1991-10-01

    Sandia National Laboratory (SNL) has proposed a new nuclear thermal propulsion concept that uses fission fragments to directly heat the propellant up to 1000 K or higher above the material temperatures. The concept offers significant advantages over traditional solid core nuclear rocket concepts because of higher propellant exit temperatures while at the same time providing for more reliable operation due to lower structure temperatures and lower power densities. The concept can be operated in either steady state or pulsed modes. The engine consists of tubular modules, each with its own pressure boundary and rocket nozzle. The steady state mode requires a large engine with a reflector for criticality, provides high thrust and high ISP. The pulse mode utilizes a driver reactor for criticality and can be considerably smaller with lower but scaleable thrust. The pulse mode does require an external heat radiator for reactor cooling, which limits its duty cycle.

  12. Fission Product Fast Reactor Constants System of JNDC

    International Nuclear Information System (INIS)

    The Fission Product Fast Reactor Constants System of JNDC has been developed for providing the FP group constants set rather automatically from the Japanese Evaluated Nuclear Data Library (JENDL). In the present version, the evaluation by JNDC was adopted for the 28 important nuclides and the evaluation by Cook was supplementally used for the other nuclides to obtain the lumped group constants. The burn-up time dependence of the lumped constants were examined. The change of capture cross sections are about 5% between 60 days and 720 days of burn-up for any type of fast reactors. The 28 important nuclides take more than 80% of total capture by fission products and cover 40% of elastic scattering and 60% of inelastic scattering. The JNDC FP lumped constants were compared with those based on Cook's evaluation and on the ENDF/B-4. The discrepancies among the three are 15% for capture and 10% for both of elastic and inelastic scattering. A benchmark test was performed using the integral measurements made in RCN, Petten, the Netherlands, in order to check the reliability of the JNDC FP group constants. The JNDC constants give better agreements than the Cook and ENDF/B-4 constants with the experiments both for FP mixtures and for separated isotopes. (auth.)

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

    International Nuclear Information System (INIS)

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

  14. Uncertainties analysis of fission fraction for reactor antineutrino experiments using DRAGON

    CERN Document Server

    Ma, X B; Chen, Y X; Zhong, W L; An, F P

    2014-01-01

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulation to predict reactor rates. First, DRAGON was developed to calculate the fission rates of the four most important isotopes in fissions,235U,238U,239Pu and141Pu, and it was validated for PWRs using the Takahama benchmark. The fission fraction calculation function was validated through comparing our calculation results with MIT's results. we calculate the fission fraction of the Daya Bay reactor core, and compare its with those calculated by the commercial reactor simulation program SCIENCE, which is used by the Daya Bay nuclear power plant, and the results was consist with each other. The uncertainty of the antineutrino flux by the fission fraction was studied, and the uncertainty of the antineutrino flux by the fission fraction simulation is 0.6% per core for Daya Bay antineutrino experiment.

  15. Coil Design and Related Studies for the Fusion-Fission Reactor Concept SFLM Hybrid

    OpenAIRE

    Hagnestål, Anders

    2012-01-01

    A fusion-fission (hybrid) reactor is a combination of a fusion device and a subcritical fission reactor, where the fusion device acts as a neutron source and the power is mainly produced in the fission core. Hybrid reactors may be suitable for transmutation of transuranic isotopes in the spent nuclear fuel, due to the safety margin on criticality imposed by the subcritical fission core. The SFLM Hybrid project is a theoretical project that aims to point out the possibilities with steady-state...

  16. Charged particle-induced nuclear fission reactions – Progress and prospects

    Indian Academy of Sciences (India)

    S Kailas; K Mahata

    2014-12-01

    The nuclear fission phenomenon continues to be an enigma, even after nearly 75 years of its discovery. Considerable progress has been made towards understanding the fission process. Both light projectiles and heavy ions have been employed to investigate nuclear fission. An extensive database of the properties of fissionable nuclei has been generated. The theoretical developments to describe the fission phenomenon have kept pace with the progress in the corresponding experimental measurements. As the fission process initiated by the neutrons has been well documented, the present article will be restricted to charged particle-induced fission reactions. The progress made in recent years and the prospects in the area of nuclear fission research will be the focus of this review.

  17. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    Science.gov (United States)

    Žerovnik, Gašper; Kaiba, Tanja; Radulović, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. PMID:25479432

  18. Microscopic modeling of mass and charge distributions in the spontaneous fission of 240Pu

    CERN Document Server

    Sadhukhan, Jhilam; Schunck, Nicolas

    2016-01-01

    In this letter, we outline a methodology to calculate microscopically mass and charge distributions of spontaneous fission yields. We combine the multi-dimensional minimization of collective action for fission with stochastic Langevin dynamics to track the relevant fission paths from the ground-state configuration up to scission. The nuclear potential energy and collective inertia governing the tunneling motion are obtained with nuclear density functional theory in the collective space of shape deformations and pairing. We obtain a quantitative agreement with experimental data and find that both the charge and mass distributions in the spontaneous fission of 240Pu are sensitive both to the dissipation in collective motion and to adiabatic characteristics.

  19. The Oarai Branch of IMR, Tohoku University as open facility for university researchers utilizing fission reactors

    International Nuclear Information System (INIS)

    For advanced future research activities utilizing fission reactors and hot laboratories, effective interlinks among fission reactors and hot laboratories are indispensable. Oarai Branch of Institute for Materials Research in Tohoku University has been playing an important role for supplying related tools for university researchers, in fission reactor irradiation and post irradiation examinations, under tight collaboration with JAERI and JNC. Now the Oarai Branch is planning to expand its collaborative functions, utilizing multi-reactors over the world and making effective interlinks among related hot laboratories in several institutions. The talk will give rough view of the present plan of the Oarai Branch, IMR, Tohoku University for tight and effective collaboration among institutions. (author)

  20. Evolution of nuclear fission reactors: Third generation and beyond

    International Nuclear Information System (INIS)

    Nuclear energy is attracting new interest around the world as countries look for low-carbon alternatives to fossil fuels to increase the diversity of their sources of energy and improve security of supply. Nuclear fission reactors provided approximately one sixth of the world's electricity needs in recent years. The vast majority of these reactors were built in the seventies and eighties. They are thus considered second generation systems, as they are based on experience gained with the first generation or prototypes built in the fifties and early sixties. Third generation reactors, developed in the nineties, are already a reality and will dominate the market in the coming decades. A significant research effort is underway on systems of the fourth generation. Better economics, improved use of natural resources, less production of radioactive waste, competitive production of hydrogen, and increased resistance to proliferation are within reach with these new systems. A review will be done on the most important features of third and fourth generation systems, together with a brief overview of the R and D challenges to be met.

  1. Beta decay of fission products for the non-proliferation and decay heat of nuclear reactors

    International Nuclear Information System (INIS)

    Today, nuclear energy represents a non-negligible part of the global energy market, most likely a rolling wheel to grow in the coming decades. Reactors of the future must face the criteria including additional economic but also safety, non-proliferation, optimized fuel management and responsible management of nuclear waste. In the framework of this thesis, studies on non-proliferation of nuclear weapons are discussed in the context of research and development of a new potential tool for monitoring nuclear reactors, the detection of reactor antineutrinos, because the properties of these particles may be of interest for the International Agency of Atomic Energy (IAEA), in charge of the verification of the compliance by States with their safeguards obligations as well as on matters relating to international peace and security. The IAEA encouraged its member states to carry on a feasibility study. A first study of non-proliferation is performed with a simulation, using a proliferating scenario with a CANDU reactor and the associated antineutrinos emission. We derive a prediction of the sensitivity of an antineutrino detector of modest size for the purpose of the diversion of a significant amount of plutonium. A second study was realized as part of the Nucifer project, an antineutrino detector placed nearby the OSIRIS research reactor. The Nucifer antineutrino detector is dedicated to non-proliferation with an optimized efficiency, designed to be a demonstrator for the IAEA. The simulation of the OSIRIS reactor is developed here for calculating the emission of antineutrinos which will be compared with the data measured by the detector and also for characterizing the level of background noises emitted by the reactor detected in Nucifer. In general, the reactor antineutrinos are emitted during radioactive decay of fission products. These radioactive decays are also the cause of the decay heat emitted after the shutdown of a nuclear reactor of which the estimation is an

  2. Method of Fission Product Beta Spectra Measurements for Predicting Reactor Anti-neutrino Emission

    OpenAIRE

    Asner, D. M.; Burns, K; Campbell, L. W.; Greenfield, B.; Kos, M. S.; Orrell, J. L.; Schram, M.; VanDevender, B.; Wood, L. S.; Wootan, D. W.

    2014-01-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron rich fission products that subsequently beta decay and emit electron anti-neutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to current precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurem...

  3. EDB-II validated, key fission product yields for fast reactor application

    International Nuclear Information System (INIS)

    Relative fission yields were measured for three different locations in the row 4 ''Test Region'' of the EBR-II reactor. Correlation of the relative fission yields to the measured average energy (anti E) and the measured 137Cs 238U/235U spectral indices have been made. The measured relative fission yields for selected fission products from 235U, 238U, 239Pu and 237Np have been compared with those values reported by the Interlaboratory Reaction Rate (ILRR) program, EBR-II fast reactor yields from destructive analysis and summation, and the March 1977 version of ENDF/B-V

  4. Dependence of Fission-Fragment Primary Charge on Nuclear Structure

    International Nuclear Information System (INIS)

    Assuming a quasi-static scission configuration the potential energy of this configuration has been calculated. The energy release between saddle point and scission point has been maximized using the liquid drop model and taking into account the mass dependence of the deform ability of the fission fragments. The leading terms that determine the charge distribution depend on the Coulomb and the asymmetry energy of the scission configuration. The deformability of the fragments shows dependence from nuclear structure. A term proportional to the difference of the deformation energies of the two fragments gives rise to a strong influence from this nuclear structure effect. In the region of closed shells the difference in the deformation energies is much the same as the total deformation energy. A term comparable to the Coulomb energy term has to be taken into account in these mass regions. The calculated Zp-values have been compared to experimental results for U235. Calculated Zp-values for Cf252 are given. (author)

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

    International Nuclear Information System (INIS)

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

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

    CERN Document Server

    Kessler, Günter

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

  8. Fission product release out of the core of a pebble bed reactor in core heatup accidents

    International Nuclear Information System (INIS)

    This report presents the analysis of fission product release from the core of a pebble-bed high temperature reactor during hypothetical accidents. First the models describing fission product transport are discussed, and on the basis of these models a computer code is developped. This code includes the diffusion of fission products from particles and through the graphite, and the sorption of metallic fission product elements on graphite as well as the plateout of metallic fission product elements in the top- and bottom reflectors. In addition a review of the necessary empirical input data is given. Then the cesium release of a single fuel element at high temperatures is calculated, and the results are compared with experimental data. Furthermore calculations of the fission product release from the core of a 500 MW(th) high temperature reactor during core heatup accidents are made, and the influence of the most important parameters is described. (orig.)

  9. Coupling of mass and charge distributions for low excited nuclear fission

    International Nuclear Information System (INIS)

    The simple model for calculation of charge distributions of fission fragments for low exited nuclear fission from experimental mass distributions is offered. The model contains two parameters, determining amplitude of even-odd effect of charge distributions and its dependence on excitation energy. Results for reactions 233U(nth,f), 235U(nth,f), 229Th(nth,f), 249Cf(nth,f) are spent

  10. On the feasibility of a fusion-fission hybrid reactor driven by dense magnetized plasmas

    International Nuclear Information System (INIS)

    The feasibility of a fusion-fission hybrid reactor driven by dense magnetized plasmas was analyzed from the point of view of the technical requirements for the fusion and fission components of the reactor. In the conceptual design, a 200 MW hybrid fusion-fission reactor is considered to be used as a heat source for district heating. The fission heat-generating blanket is based on the CANDU reactor technology, while the fusion fast neutrons are provided by a high-density pinch plasma. As far as the fission components of the reactor are concerned, the hybrid reactor turns out to be entirely feasible based on existing technologies. On the other hand extensive development will be needed to meet the requirements for the fusion component of the reactor. The basic conditions for a dense magnetized plasma fusion device to be used for the proposed hybrid concept are not concerned only with the attainment of high neutron yield per pulse (at least 5 x 10 18), but also with a relatively high repetition rate (in the range 1-10 Hz). An important feature of the proposed design is its inherent safety feature: no active component are necessary within the reactor containment area, all the hybrid system control being ensured by the fusion component of the reactor. (authors)

  11. Fuels and fission products clean up for molten salt reactor of the incinerator type

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Gorbunov, V.; Zakirov, R. [RRC-Karchatov Institute, Moscow (Russian Federation)

    2000-07-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  12. Fuels and fission products clean up for molten salt reactor of the incinerator type

    International Nuclear Information System (INIS)

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  13. Automated training system simulating charging machine operation for repair personnel education at NPPs with WWER-1000

    International Nuclear Information System (INIS)

    An automated training system (ATS) used as a full-scale simulator of real fuel charging machine (CM) for the WWER-1000 reactor in the process of repair personnel education is described. The main ATS specific features are the presence of one, but not three monitors and accelerated operation regime. CM consol is imaged on PC display and images of three pseudomonitors are simulated in the following combinations: alphanumerical and TV monitors; alphanumerical and graphic ones

  14. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  15. Characteristics of light charged particle emission in spontaneous and neutron induced fission of Cm and Cf isotopes

    OpenAIRE

    Vermote, Sofie

    2009-01-01

    Talking about fission, people think almost automatically about a nucleus that splits into two heavy pieces. However, in 1946 the phenomenon of ternary fission has been discovered, in which the two fission fragments are accompanied by a light charged particle. Ternary fission data are of interest for nuclear physics, however also the nuclear industry requests accurate data for ternary fission yields, especially 3H and 4He, since they are at the origin of the production of He gas and the radioa...

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

    International Nuclear Information System (INIS)

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

  17. A new MC-based method to evaluate the fission fraction uncertainty at reactor neutrino experiment

    OpenAIRE

    Ma, X. B.; Qiu, R. M.; Y. X. Chen

    2016-01-01

    Uncertainties of fission fraction is an important uncertainty source for the antineutrino flux prediction in a reactor antineutrino experiment. A new MC-based method of evaluating the covariance coefficients between isotopes was proposed. It was found that the covariance coefficients will varying with reactor burnup and which may change from positive to negative because of fissioning balance effect, for example, the covariance coefficient between $^{235}$U and $^{239}$Pu changes from 0.15 to ...

  18. Most probable charge of fission products in proton-induced fission of sup 2 sup 3 sup 8 U and sup 2 sup 3 sup 2 Th

    CERN Document Server

    Kaji, D; Kudo, H; Fujita, M; Shinozuka, T; Fujioka, M

    2002-01-01

    The charge distributions of fission products in proton-induced fission of sup 2 sup 3 sup 8 U and sup 2 sup 3 sup 2 Th were measured in a wide mass range. The most probable charges lay on the proton-rich side in the light fragment region and on the proton-deficient side in the heavy one compared with the unchanged charge distribution hypothesis. This result implies that the charge polarization occurs in the fission process. The charge polarization was examined with respect to the ground-state Q values. The estimations by the Q values fairly well reproduced the experimental most probable charges. These results suggest that the fission path to the most favorable charge division may go through the most energetically favorable path at scission point. (author)

  19. Assessment of fission product yields data needs in nuclear reactor applications

    International Nuclear Information System (INIS)

    Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Asner, David M.; Burns, Kimberly A.; Campbell, Luke W.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wood, Lynn S.; Wootan, David W.

    2015-03-01

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

  3. Experimental study of neutrino oscillations at a fission reactor

    International Nuclear Information System (INIS)

    The energy spectrum of neutrinos from a fission reactor was studied with the aim of gaining information on neutrino oscillations. The well shielded detector was set up at a fixed position of 8.76 m from the point-like core of the Laue-Langevin reactor in an antineutrino flux of 9.8 x 1011cm-2s-1. The target protons in the reaction antiνsub(e)p → e+n were provided by liquid scintillation counters (total volume of 377l) which also served as positron detectors. The product neutrons moderated in the scintillator were detected by 3He wire chambers. A coincidence signature was required between the prompt positron and the delayed neutron events. The positron energy resolution was 18% FWHM at 0.91 MeV. The signal-to-background ratio was better than one to one between 2 MeV and 6 MeV positron energy. At a counting rate of 1.58 counts per hour, 4890+-180 neutrino induced events were detected. The shape of the measured positron spectrum was analyzed in terms of the parameters Δ2 and sin2 2theta for two-neutrino oscillations. The experimental data are consistent with no oscillations. An upper limit of 0.15 eV2 (90% c.l.) for the mass-squared differences Δ2 of the neutrinos was obtained, assuming maximum mixing of the two neutrino states. The ratio of the measured to the expected integral yield of positrons assuming no oscillations was determined to be ∫Ysub(exp)/∫Ysub(th) = 0.955+-0.035 (statistical)+-0.110 (systematic)

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

    International Nuclear Information System (INIS)

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

  5. A fission fragment reactor concept for nuclear thermal propulsion

    Science.gov (United States)

    Suo-Anttila, Ahti J.; Parma, Edward J.; Pickard, Paul S.; Wright, Steven A.; Vernon, Milton E.

    1992-01-01

    The Space Exploration Initiative requires the development of nuclear thermal and nuclear electric technologies for space propulsion for future Luna and Mars missions. Sandia National Laboratories has proposed a new nuclear thermal propulsion concept that uses fission fragments to directly heat the propellant up to 1000 K or higher above the material temperatures. The concept offers significant advantages over traditional solid-core nuclear rocket concepts because of higher propellent exit temperatures, while at the same time providing for more reliable operation due to lower structure temperatures and lower power densities. The reactor can be operated in either a steady-state or pulsed mode. The steady-state mode provides a high thrust and relatively high specific impulse, as compared to other nuclear thermal concepts. The pulsed mode requires an auxillary radiator for cooling, but has the possibility of achieving very high specific impulses and thrust scaleable to the radiator size. The propellant temperatures are limited only by thermal radiation and transient heat conduction back to the substrate walls.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-10-12

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

  8. Temperature transients of a fusion-fission ITER pebble bed reactor in loss of coolant accident

    International Nuclear Information System (INIS)

    In this preliminary scoping study, post-accident temperature transients of several fusion-fission designs utilizing ITER-FEAT-like parameters and fission pebble bed fuel technology are examined using a 1-D cylindrical MATLAB heat transfer code along with conventional fission decay heat approximations. Scenarios studied include systems with no additional passive safety features to systems with melting reflectors designed to increase emissivity after reaching a specified temperature. Results show that for a total fission power of ∼1400-2800 MW, two of the realistic variants investigated are passively safe. The crucial time, defined as the time when either any structural part of the fusion-fission tokamak reaches melting point, or when the pebble fuel reaches 1873 K, ranges from 5.7 to 76 h for the unsafe configurations. Additionally, it is illustrated that, fundamentally, the LOCA characteristics of pure fission pebble beds and fusion-fission pebble beds are different. Namely, the former depends on the pebble fuel's large thermal capacity, along with external radiation and natural convective cooling, while the latter depends significantly more on the tokamak's sizeable total internal heat capacity. This difference originates from the fusion-fission reactor's conflicting goal of having to minimize heat transfer to the magnets during normal operation. These results are discussed in the context of overall fusion-fission reactor design and safety

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  10. Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

    International Nuclear Information System (INIS)

    The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations

  11. A new MC-based method to evaluate the fission fraction uncertainty at reactor neutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X

    2016-01-01

    Uncertainties of fission fraction is an important uncertainty source for the antineutrino flux prediction in a reactor antineutrino experiment. A new MC-based method of evaluating the covariance coefficients between isotopes was proposed. It was found that the covariance coefficients will varying with reactor burnup and which may change from positive to negative because of fissioning balance effect, for example, the covariance coefficient between $^{235}$U and $^{239}$Pu changes from 0.15 to -0.13. Using the equation between fission fraction and atomic density, the consistent of uncertainty of fission fraction and the covariance matrix were obtained. The antineutrino flux uncertainty is 0.55\\% which does not vary with reactor burnup, and the new value is about 8.3\\% smaller.

  12. Sustainable and safe nuclear fission energy. Technology and safety of fast and thermal nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, Guenter

    2012-07-01

    Written by one of the world-leading specialists in reactor physics and safety Most comprehensive book on nuclear fission technology, new safety concepts and waste disposal Complete description and evaluation of nuclear fission power generation Covers the whole nuclear fuel cycle, from the extraction of natural uranium, uranium conversion and enrichment up to the fabrication of fuel elements Description of the different fuel cycle options Presents viable solutions for safe and long-term storage of nuclear waste Recently developed new safety concepts for fission reactors Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  13. Super-heavy nuclei with Z = 118 and their mass and charge spectrum of fission fragments

    Science.gov (United States)

    Maslyuk, V. T.; Smolyanyuk, A. V.

    2015-12-01

    The first results of the calculation of the mass and charge yields of fission fragments for over 60 isotopes which have Z = 118 are presented. The results were obtained from the condition of thermodynamic ordering of the ensemble of fission fragments. The role of neutrons shells with N = 82 or N = 126 and protons shells with Z = 50 in the realization of symmetric (or one-humped) and asymmetric (2- or 3-humped) shapes of the fission-fragment yields with the transition from neutron-proficient to neutron-deficient isotopes was investigated. The data of fragments yields had been analyzed under the conditions of a “cold” and “hot” fission. The calculations show the possibility to identify super-heavy nuclei with Z ≥ 118 produced synthetically by heavy-ion reaction on their mass/charge spectrum division.

  14. Quantification of structural materials for reactor systems: synergy's in materials for fusion/fission reactors and advanced fission reactor

    International Nuclear Information System (INIS)

    In nuclear technology a lot of experience has been accumulated meanwhile from reactor programmes for ferritic alloys, austenitic steels and Ni-based alloys as main component materials during R and D, design, construction and operation. Generally materials are a key issue for a safe and reliable operation of -NPPs. Many grades investigated are of interest for the design of GenIVs and fusion reactors. Synergisms of materials, material technologies, mechanical data, corrosion and other topics -for the qualification of materials for nuclear systems are generally discussed and information on a qualification procedure is compiled. Also some lessons learned from fabrication, test programmes or operation of NPPs are provided. A special problem is the fusion system because a final validation for alloy performance in the long term will need irradiation under realistic -fusion condition anticipated in a high-energetic, fusion-specific intense neutron source such as (IFMIF), the International Fusion Materials Irradiation Facility. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L. (ed.)

    1979-07-01

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

  17. High temperature and sensitivity fission chambers: qualification of the CFUCO7 in reactor

    International Nuclear Information System (INIS)

    We present, in this paper, the whole tests performed both in laboratory and in reactor on the high temperature, wide dynamic fission chamber CFUCO7 and on its associated electronics. Except the long time tests to be realized in the PHENIX reactor, this measurement device, fission chamber and wide range electronic, can be considered as qualified to be used in a large LMFBR. We present also the new improvements on the detector design and the future programme in the reactor SUPER-PHENIX. (authors). 9 figs., 4 tabs., 2 refs., 2 appendix

  18. Fault Diagnosis of Batch Reactor Using Machine Learning Methods

    OpenAIRE

    2014-01-01

    Fault diagnosis of a batch reactor gives the early detection of fault and minimizes the risk of thermal runaway. It provides superior performance and helps to improve safety and consistency. It has become more vital in this technical era. In this paper, support vector machine (SVM) is used to estimate the heat release (Qr) of the batch reactor both normal and faulty conditions. The signature of the residual, which is obtained from the difference between nominal and estimated faulty Qr values,...

  19. Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications

    CERN Document Server

    Herndon, J Marvin

    2013-01-01

    The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent ...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  2. Fission fragment charge and mass distributions in 239Pu(n,f) in the adiabatic nuclear energy density functional theory

    OpenAIRE

    Regnier, D.; Dubray, N.; Schunck, N.; Verriere, M.

    2016-01-01

    Accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r-process to fuel cycle optimization for nuclear energy. The need for a predictive theory applicable where no data is available is an incentive to develop a fully microscopic approach to fission dynamics. In this work, we calculate the pre-neutron emission charge and mass distributions of the fission fragments formed in the neutron-induced fission of ...

  3. Charge and mass distribution in 7Li induced fission of 232Th

    International Nuclear Information System (INIS)

    Formation cross sections of about forty fission products have been determined using recoil catcher technique followed by off line gamma-ray spectrometry in 7Li induced fission of 232Th at Elab=41.9, 36.6 and 31.4 MeV. The measured data have been used to deduce charge and mass distributions. Mass distribution is found to be asymmetric at all the three energies. Cross sections of evaporation residues formed in both transfer reactions (232,233,234Pa) as well as in complete fusion (234Np), have also been measured. The measured evaporation residue cross sections and the decay probabilities of target like nuclei (233,234,235Pa) formed in the various transfer reactions, as calculated by PACE2, have been used to estimate the transfer induced fission cross sections. The data indicated that the magnitude of transfer induced fission is very small

  4. Investigations on radioactive and stable fission gas release behaviour at the Halden reactor

    International Nuclear Information System (INIS)

    Two types of experiments have been used in the Halden reactor to investigate the release of fission gases from LWR fuel. The first employs internal pressure sensors from which the kinetics and quantity of stable gases can be measured during irradiation. The second is the use of sweep gases to carry released fission gases from the fuel rod to a detector situated outside the reactor. With this equipment, it is possible to measure, using gamma spectroscopy, both radioactive and stable fission product release. In conjunction with fuel centerline thermocouples to measure fuel temperatures, these techniques have been successful in improving our understanding of the release process and the factors affecting it. The data generated have been used in many member countries to develop models and validate fuel performance codes used in reactor safety assessments. (authors)

  5. Nuclear charge distribution in the spontaneous fission of 252Cf

    OpenAIRE

    Wang, Taofeng; Zhu, Liping; WANG, LIMING; Men, Qinghua; Han, Hongyin; Xia, Haihong

    2015-01-01

    The measurement for charge distributions of fragments in 252Cf has been performed by using a unique style of detector setup consisting of a typical grid ionization chamber and a dE-E particle telescope. We found that the fragment mass dependency of the average width of the charge distribution shows a systematic decreased trend with the obvious odd-even effect. The variation of widths of charge distribution with kinetic energies shows an approximate V-shape curve due to the large number of neu...

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

    International Nuclear Information System (INIS)

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

  7. Dimensional control and check of field machining parts for reactor internals installation

    International Nuclear Information System (INIS)

    Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

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

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

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

  9. Fission product filter for hot reactor cooling gas

    International Nuclear Information System (INIS)

    The fission product filter for He consists of a winding body composed of two corrugated metal sheets simultaneously wound on a core laterally reversed. It is inserted into an enclosing tube and held at top and bottom by a star-shaped yoke. (DG)

  10. Relative fission product yield determination in the USGS TRIGA Mark I reactor

    Science.gov (United States)

    Koehl, Michael A.

    Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular

  11. Analysis of reactor coolant system depressurization effect of ex-vessel release of fission products

    International Nuclear Information System (INIS)

    Coupling model of thermal-hydraulic, fission products behavior and radiological consequences assessment was constructed. Based on high-pressure core melt severe accidents of SB-LOCA, SGTR, SBO, and LOFW, the influence of reactor coolant system (RCS) depressurization on ex-vessel release of fission products was studied, including mitigation effect on ex-vessel release of fission products and other negative effects. It is shown that RCS depressurization can mitigate ex-vessel release of fission products for high-pressure core melt accident sequences, while airborne activity in the early phase with RCS depressurization is higher than that of the base cases without RCS depressurization. The research results can give support to establish severe accident management guideline. (authors)

  12. A model for non-volatile fission product release during reactor accident conditions

    International Nuclear Information System (INIS)

    An analytical model has been developed to describe the release kinetics of non-volatile fission products (e.g., Mo, Ce, Ru and Ba) from uranium dioxide fuel under severe reactor accident conditions. The present treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of Sb and Te in the Zircaloy cladding. This model has been used to interpret the release behaviour of fission products observed in the CEA experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere. (author)

  13. Support vector machines for nuclear reactor state estimation

    International Nuclear Information System (INIS)

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

  14. Determination of the Primary Nuclear Charge of Fission Fragments from their Characteristic K-X-Ray Emission in Spontaneous Fission of Cf252

    International Nuclear Information System (INIS)

    The distribution of nuclear charge in the spontaneous fission of Cf252 has been determined directly by simultaneous measurement of the masses and characteristic K-X-ray energies associated with the primary fission products. The X-rays were detected by a thin Nal (Tl) crystal (or by an argon-filled proportional counter) in coincidence with a pair of solid-state detectors for the complementary fission fragments. Preliminary to the three-parameter study of the charge-mass distribution the gross characteristics of the K-X-rays were examined in some detail. The average yield of K-X-rays is 0.55 ± 0.1 pet fission (the heavy group accounting fot 70% of the total). From delayed-coincidence and fragment time-of-flight experiments it was.found that about 30% of the X-rays are emitted within 0.1 ns after fission, another 30% between 0.1 and 1 ns, 25% between 1 and 10 ns, the remainder appearing as two delayed components of equal intensity with half-lives of ∼30 ns and ∼100 ns. These characteristics indicate that the X-rays arise from internal conversion during de-excitation of the primary fission fragments, an interpretation supported by the observed yield 1 per fission) of 50 - 300 - keV electrons emitted within 2 ps of fission. In the three-parameter experiments the yield and energy of K-X-rays emitted in the first centimeter (ns) of fragment flight were determined as a function of fragment mass. The yield of K-X-rays per fragment is a pronounced saw-tooth function of mass, rising from p) function in better agreement with the empirical rule of equal charge displacement (ECD) than with other postulates for charge division in nuclear fission. (author)

  15. Measurements of charge distributions of the fragments in the low energy fission reaction

    International Nuclear Information System (INIS)

    The measurement for charge distributions of fragments in spontaneous fission 252Cf has been performed by using a unique style of detector setup consisting of a typical grid ionization chamber and a ΔΕ−Ε particle telescope, in which a thin grid ionization chamber served as the ΔΕ-section and the E-section was an Au–Si surface barrier detector. The typical physical quantities of fragments, such as mass number and kinetic energies as well as the deposition in the gas ΔΕ detector and E detector were derived from the coincident measurement data. The charge distributions of the light fragments for the fixed mass number A2⁎ and total kinetic energy (TKE) were obtained by the least-squares fits for the response functions of the ΔΕ detector with multi-Gaussian functions representing the different elements. The results of the charge distributions for some typical fragments are shown in this article which indicates that this detection setup has the charge distribution capability of Ζ:ΔΖ>40:1. The experimental method developed in this work for determining the charge distributions of fragments is expected to be employed in the neutron induced fissions of 232Th and 238U or other low energy fission reactions.

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

    International Nuclear Information System (INIS)

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

  17. Isotope production in light charged particle induced fission

    International Nuclear Information System (INIS)

    In this thesis the production of neutron-rich isotopes in the mass region A=96-120 has been studied. Their yields have been extensively studied with the IGISOL isotope separator in the 238U (p,f) reaction. Deuteron and alpha particle induced reactions have also been investigated. In connection with this work several new isotopes have been identified for the first time. Specifically, the decays of 110Tc and 112Tc are discussed. Cumulative mass distributions as well as independent isotopic distributions have been constructed. In the method used here the total kinetic energy of the fragments is integrated so that there is no energy selection and the yields are post-neutron emission values. A theoretical model described in one of the joined papers has been used to extract the preneutron emission yields. These results can be used for estimating the possibilities of production of new neutron-rich isotopes, e.g. at the IGISOL facility. In addition, they are necessary to help refining the existing fission models. (orig.) (50 refs., 9 figs.)

  18. Markets for reactor-produced non-fission radioisotopes

    International Nuclear Information System (INIS)

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

  19. Human machine interface for research reactor instrumentation and control system

    International Nuclear Information System (INIS)

    Most present design of Human Machine Interface for Research Reactor Instrumentation and Control System is modular-based, comprise of several cabinets such as Reactor Protection System, Control Console, Information Console as well as Communication Console. The safety, engineering and human factor will be concerned for the design. Redundancy and separation of signal and power supply are the main factor for safety consideration. The design of Operator Interface absolutely takes consideration of human and environmental factors. Physical parameters, experiences, trainability and long-established habit patterns are very important for user interface, instead of the Aesthetic and Operator-Interface Geometry. Physical design for New Instrumentation and Control System of RTP are proposed base on the state-of- the-art Human Machine Interface design. (author)

  20. Nuclear fission reactors from thousand of million years; Reactores de fision nuclear de hace miles de millones de anos

    Energy Technology Data Exchange (ETDEWEB)

    Bulbulian G, S.; Ordonez R, E.; Fernandez V, S.M. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2005-07-01

    This book is about nuclear reactors, not only of the industrial ones that work to provide electric power, neither of those experimental ones as the first one that worked in Chicago in the first half of the XX Century but, mainly, of those that worked in the Earth thousands of millions of years ago. The book examines what happened in last geologic times, when the natural uranium had a different constitution to the current one. We will give you information on the nuclear fission reactors that worked in Gabon, Africa. A discussion of the radioactive isotopes formed during the operation of those reactors and its behavior until our days is presented. (Author)

  1. Effect of fission products accumulation on thermophysical properties of oxide fuels for fast reactors

    International Nuclear Information System (INIS)

    It is important to understand the behavior of fission products under irradiation. In this paper, recent activities for obtaining the fundamental findings concerning the effect of FPs accumulation on the thermophysical properties of oxide fuels for fast reactors are presented. (author)

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

    International Nuclear Information System (INIS)

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

  3. Use of fission reactors for the generation of electricity in the UK civil programme

    International Nuclear Information System (INIS)

    This is a statement issued by the Institution of Nuclear Engineers, London, U.K., with the object of informing the public of the basic facts concerning nuclear power, including ethics and safety. It concentrates on the use of fission reactors and considers society's needs in the immediate future. (U.K.)

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

    International Nuclear Information System (INIS)

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

  5. Theory of fission detector signals in reactor measurements

    CERN Document Server

    Pál, L

    2015-01-01

    The Campbell theorem, relating the variance of the current of a fission chamber (a "filtered Poisson process") to the intensity of the detection events and to the detector pulse shape, becomes invalid when the neutrons generating the fission chamber current are not independent. Recently a formalism was developed by the present authors [1], by which the variance of the detector current could be calculated for detecting neutrons in a subcritical multiplying system, where the detection events are obviously not independent. In the present paper, the previous formalism, which only accounted for prompt neutrons, is generalised to account also for delayed neutrons. A rigorous probabilistic analysis of the detector current was performed by using the same simple, but realistic detector model as in the previous work. The results of the present analysis made it possible to determine the bias of the traditional Campbelling techniques both qualitatively and quantitatively. The results show that the variance still remains ...

  6. Machine learning of the reactor core loading pattern critical parameters

    International Nuclear Information System (INIS)

    The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employed a recently introduced machine learning technique, Support Vector Regression (SVR), which has a strong theoretical background in statistical learning theory. Superior empirical performance of the method has been reported on difficult regression problems in different fields of science and technology. SVR is a data driven, kernel based, nonlinear modelling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modelling. The starting set of experimental data for training and testing of the machine learning algorithm was obtained using a two-dimensional diffusion theory reactor physics computer code. We illustrate the performance of the solution and discuss its applicability, i.e., complexity, speed and accuracy, with a projection to a more realistic scenario involving machine learning from the results of more accurate and time consuming three-dimensional core modelling code. (author)

  7. Fission fragment charge and mass distributions in 239Pu(n,f) in the adiabatic nuclear energy density functional theory

    CERN Document Server

    Regnier, D; Schunck, N; Verriere, M

    2016-01-01

    Accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r-process to fuel cycle optimization for nuclear energy. The need for a predictive theory applicable where no data is available is an incentive to develop a fully microscopic approach to fission dynamics. In this work, we calculate the pre-neutron emission charge and mass distributions of the fission fragments formed in the neutron-induced fission of 239Pu using a microscopic method based on nuclear energy density functional (EDF) method, where large amplitude collective motion is treated adiabatically using the time dependent generator coordinate method (TDGCM) under the Gaussian overlap approximation (GOA). Fission fragment distributions are extracted from the flux of the collective wave packet through the scission line. We find that the main characteristics of the fission charge and mass distributions can be well reproduced by existing energy functionals even in tw...

  8. Fission product release in accidents in light water reactors

    International Nuclear Information System (INIS)

    The author deals with the three phases of release from the reactor core, from the reactor system, and finally from the containment. Particular interest is given to the release from the reactor core at temperatures which let the fuel rod cladding burst leading to meltdown of the fuel elements and evaporation from the core melt. The special case of the steam explosion with small nuclear fuel particles pouring out into an oxidating atmosphere is touched upon. The Rasmussen study is the basis of the statements. (orig./LH)

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    WANG Xin-Hua; GUO Hai-Ping; MOU Yun-Feng; ZHENG Pu; LIU Rong; YANG Xiao-Fei; YANG Jian

    2013-01-01

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

  11. Reactor inspection and maintenance machine senses and homes in on reactor end fittings

    International Nuclear Information System (INIS)

    The Universal Delivery Machine (UDM) is a new CANDU reactor maintenance tool that allows safe, timely, and cost-effective inspection and maintenance of fuel channels. The UDM must align precisely with reactor end-fittings in order to clamp onto fuel channels without applying excessive force. This alignment process is called fine homing. This paper describes the UDM instrumentation and control design features used in the fine homing process. (author)

  12. Survey of experimental studies on release and deposition of reactor fuel fission products

    International Nuclear Information System (INIS)

    In the work, a review of the most important results from 11 series of experimental studies on fission product behaviour in reactor accident conditions that were performed in a number of research facilities worldwide, is presented. The facilities can be divided into out-of-pile, in-pile and integral ones, the later ones modelling the whole of a reactor cooling system. Emphasis is given not only on quantitative description of release and deposition phenomena but on physico-chemical processes accompanying radioactivity migration in reactor circuits and variety of FP chemical forms as well. (author)

  13. Nuclear charge distribution of fission products originated from fission of 238U nuclei induced by45-69 MeV protons

    Directory of Open Access Journals (Sweden)

    Houshyar Noshad

    2007-12-01

    Full Text Available  Fission of 238U nuclei was performed by 45-69 MeV protons at the Cyclotron and Radioisotope Center of Tohoku University in Japan. The fission products originated in the reaction were identified by using gamma spectroscopy. The experimental data show that the charge distribution of isobar fission products follows a Gaussian distribution with a standard deviation independent of the selected mass number. The standard deviations were measured for the reaction 238U(p, f with 45, 55, 65 and 69 MeV protons. For Ep = 45 MeV, the standard deviation obtained from the experiment is in agreement with the existing data and satisfies the prediction of the Hauser-Feshbach statistical model. For other proton energies, measurement of this quantity has not been reported in the literature. The experimental results show that the value of standard deviation increases, when the excitation energy of the fissioning nucleus increases. Furthermore, the most probable charge was determined for the isobar fission products detected in the experiment. The results are consistent with the prediction of the minimum potential energy (MPE model. Moreover, the experimental data show that nuclear charge polarization occurs in the fission process.

  14. Fission product transport in the reactor coolant system for a spectrum of interfacing system LOCA scenarios

    International Nuclear Information System (INIS)

    One of the most important potential severe accident sequences for any pressurized water reactor (PWR) is a loss of coolant accident (LOCA), or V-sequence, in one of the interfacing systems. As initially described in the reactor safety study WASH-1400, interfacing system LOCAs involved the failure of check valves in emergency core cooling systems (ECCS), but could also involve the residual heat removal (RHR) systems. The check valves protect the low-pressure portions of these systems from the high pressures of the reactor coolant system (RCS) to which they are connected to provide cold leg injection. A consequent break in the low-pressure piping outside the containment may result in core damage and a direct pathway for fission products to be transported from the core, through the RCS and ECCS or RHR to the auxiliary building, from which they can escape to the environment. This paper addresses the retention and transport of fission products (specifically, CsI) in the RCS in V-sequence scenarios. It summarizes some of the major differences between models resulting from the latest version of the industry degraded core rulemaking (IDCOR) MAAP Computer Program, MAAP 3.0B. Discussed are the differences in: fission product transport and retention in small, medium, and large ECCS pipe breaks, as well as the effect of ECCS and auxiliary feedwater (AFW) system operation and fission product retention in the various regions of the RCS as calculated by MAAP 3.0B and the STCP

  15. Study on the influence of prompt fission γ-ray and delayed γ-ray on reactor internals heating rate

    International Nuclear Information System (INIS)

    To improve the accuracy of the calculated reactor internals heating rate in the design of nuclear power plants, this paper studied the contribution of prompt fission γ to the reactor internals heating rate based on the original method of MCNP external neutron source model. The results revealed that the reactor internals heating rate increased by 9%∼38% with prompt fission γ taken into account and the internals nearer to the core had a lager increment. In addition, it is believed after analysis that the contribution of the delayed γ on reactor internals heating rate is similar to the prompt fission γ. Therefore, when calculating reactor internals heating rate, in addition to the neutron source and neutron capture γ, prompt fission γ and delayed γ should also be considered. (authors)

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

    International Nuclear Information System (INIS)

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

  17. Status of pseudo-fission-product cross-sections for fast reactors

    International Nuclear Information System (INIS)

    Within the framework of the Subgroup 17 (SG17) benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (FR), a comparison of lumped or pseudo-fission-product cross-sections for fast reactors has been made. Several parameters have been compared: the one- group cross-sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross-sections, and the one-group cross sections of individual fission products. Graphs of the multi-group cross-sections and those of capture cross-sections for 27 nuclides have also been compared. (R.P.)

  18. Design of a heatpipe-cooled Mars-surface fission reactor

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

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

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

    Science.gov (United States)

    Porta, A.; Zakari-Issoufou, A.-A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; Estienne, M.; Agramunt, J.; Äystö, J.; Bowry, M.; Briz, J. A.; Caballero-Folch, R.; Cano-Ott, D.; Cucouanes, A.; Elomaa, V.-V.; Eronen, T.; Estévez, E.; Farrelly, G. F.; Garcia, A. R.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Karvonen, P.; Kolhinen, V. S.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez-Cerdán, A. B.; Podolyák, Zs.; Penttilä, H.; Regan, P. H.; Reponen, M.; Rissanen, J.; Rubio, B.; Shiba, T.; Sonzogni, A. A.; Weber, C.

    2016-03-01

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

  1. Hydraulic stud-tensioning machines in reactor technology

    International Nuclear Information System (INIS)

    Hydraulic multiple stud tensioner (MST) for the simultaneous prestressing of all the stud bolts is make it possible to achieve highly accurate prestress levels in the highly stressed bolts holding down the top head of reactor pressure vessels. These machines can remove and replace the nuts and studs, and can rotate these components upwards and downwards, during the operation of opening and closing the reactor pressure vessel. In order to reduce the radiation exposure of the service personnel, and also to reduce the time required for this work which may lie in the critical path of the refuelling time schedule, it is desirable to achieve complete mechanisation of these machines, including remote control and remote monitoring. The devices and components required for this purpose are without precedent in machine construction with respect to their functions and to the load range involved. The reported operating experience therefore also covers some points of general interest while the data on maintenance reflect the known status of the technology. (orig.)

  2. A contribution to the study of mass-kinetic energy-nuclear charges correlations for fission fragments with the 'Cosi Fan Tutte' spectrometer

    International Nuclear Information System (INIS)

    The present work is devoted to the fragments produced in the neutrons induced fission of 235U, performed with the time of flight-energy spectrometer 'Cosi Fan Tutte' recently built at the neutron high flux reactor of the Laue-Langevin Institute at Grenoble. Mass-kinetic energy-nuclear charge correlations were measured for the light fission fragment group. Nuclear charges were identified for the first time on this spectrometer using the range difference of the fission fragments in an axial field ionisation chamber. The present results are in good agreement with the previous one obtained using the spectrometer 'Lohengrin', which proves the validity of the methods which we developed. In addition, we extend the measurements to higher kinetic energies. The structures which appear in the distributions are attributed to spherical and deformed shell effects in the nascent fragments and to even odd effects. The study of thermal neutron induced fission of 229Th, which is scarcely known, has been started. (author)

  3. Calculation of fissile nuclides and fission products inventory applied to ETRR-1 research reactor

    International Nuclear Information System (INIS)

    The study of the nuclear reactor fuel safety implies studying physical mechanical, thermal and chemical proportions of the fuel during normal operation and accident conditions. A model was developed to calculate the fissile nuclides and fission products inventory in an operating reactor. The model considers the production and removal of different radionuclides leaking into account the decay schemes of each. The mathematical formulas were treated without any approximations. A decay model was developed for the period after reactor shutdown. The amount of different nuclides was evaluated for a given cooling time. Egypt test and research reactor number 1, ETRR-1. Was chosen to apply the model. The amount of about 200 nuclides was calculated. A certain nuclides was chosen to be presented based on their poisoning ratios. Criticality calculations were carried out to investigate the criticality condition of the reactor at different operating times. 4 fig

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

    International Nuclear Information System (INIS)

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 25000C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  5. A fusion-fission reactor driven by plasma-liner impact

    International Nuclear Information System (INIS)

    It is shown that the impact of a quasi-spherical plasma liner on a spherical solid liner can produce a highly luminous source of soft X-rays. This radiation can be used for the ablation of an inner spherical liner, which can be thus accelerated to speeds above 107 cm/sec. Such a liner should be able to compress a core of fissionable material, surrounded by a D-T mantle to fission - criticality. The burst of the fission energy then ignites the D-T mantle which produces a larger burst of fusion energy. The energy liberated in such a microexplosion is estimated to be of the order of 1 GJ. An apparatus based on a symmetrical plasma-focus geometry should be able to produce the plasma liner. A reactor combining these concepts is described. (orig.)

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

    International Nuclear Information System (INIS)

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

  7. Modelling fission gas release and swelling in fast reactor fuel pins

    International Nuclear Information System (INIS)

    Investigations into the mechanisms involved in the release of fission product gases from fast reactor fuel elements have been made. A rate theory model of the homogeneous nucleation and subsequent growth of fission gas bubbles has been developed which allows the inclusion of bubbles containing in excess of 106 atoms. Various processes which influence the growth of the intragranular bubbles and hence the release to grain boundaries have been included in the model. Bubble migration and coalescence are shown to be essential in order to bring the amount of gas release into line with experimental observations. The effects of fission gas re-solution rate, gas atom diffusivity, grain size, temperature and binary nucleation coefficient have been investigated and it is concluded that the dominant parameters for gas release are temperature, grain size, re-solution rate and bubble migration and coalescence. (author)

  8. Reaction rates in blanket assemblies of a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    To validate neutronics calculation for the blanket design of fusion-fission hybrid reactor, experiments for measuring reaction rates inside two simulating assemblies are performed. Two benchmark assemblies were developed for the neutronics experiments. A D-T fusion neutron source is placed at the center of the setup. One of them consists of three layers of depleted uranium shells and two layers of polyethylene shells, and these shells are arranged alternatively. The 238U capture reaction rates are measured using depleted uranium foils and an HPGe gamma spectrometer. The fission reaction rates are measured using a fission chamber coated with depleted uranium. The other assembly consists of depleted uranium and LiH shells. The tritium production rates are measured using the lithium glass scintillation detector which is placed in the LiH region of the assembly. The measured reaction rates are compared with the calculated ones predicted using MCNP code, and C/E values are obtained. (authors)

  9. Machine Learning of the Reactor Core Loading Pattern Critical Parameters

    International Nuclear Information System (INIS)

    The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR), which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy

  10. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  11. Mobile stand for testing charging machine ram devices

    International Nuclear Information System (INIS)

    The equipment described is designed to functional testing of the charging machine (CM) end, in laboratory or NPP conditions, over the span of maintenance activities. It appears to be a portable panel that can be easily coupled to the regulation and control elements of the ram driving systems. Such an equipment occurred as necessary following the analyses of the results and technical problems issued from the technical assistance services which INR Pitesti performed for Cernavoda NPP in the period 1996-1999. The experience acquired from these works resulted in a new design and execution of the CM ram devices the characteristics of which are indicated. The equipment was certified and is now successfully utilized at INR Pitesti and Cernavoda NPP Unit 1. The mobile stand will be used in the near future for testing operations of the CM ends number 4 and 5 destined to Cernavoda NPP Unit 2, planned for year 2002

  12. Energy dependence of mass, charge, isotopic, and energy distributions in neutron-induced fission of 235U and 239Pu

    Science.gov (United States)

    Pasca, H.; Andreev, A. V.; Adamian, G. G.; Antonenko, N. V.; Kim, Y.

    2016-05-01

    The mass, charge, isotopic, and kinetic-energy distributions of fission fragments are studied within an improved scission-point statistical model in the reactions 235U+n and 239Pu+n at different energies of the incident neutron. The charge and mass distributions of the electromagnetic- and neutron-induced fission of 214,218Ra, 230,232,238U are also shown. The available experimental data are well reproduced and the energy-dependencies of the observable characteristics of fission are predicted for future experiments.

  13. Performance targets for fusion-fission (hybrid) reactors

    International Nuclear Information System (INIS)

    The potential roles for various performance levels of hybrid power plants in the United States have been estimated for various scenarios which describe the systems with which the hybrid plants must compete. Due to the uncertainty in estimating the costs of hybrid power plants, a parametric approach was chosen to identify acceptable characteristics and associated tradeoffs. Because hybrid power plants produce substantial quantities of both plutonium and electricity, a simulation model is necessary which appropriately values each of these products. This would permit design tradeoffs and realistic interactions with other generating plants, such as Liquid Metal Fast Breeder Reactors (LMFBRs) and Light Water Reactors (LWRs), to be estimated. A linear programming model nearly identical to that used for the recent LMFBR cost-benefit analysis was utilized for this purpose

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

    Science.gov (United States)

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

    2016-02-01

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

  15. Nuclear energy and fusion-fission hybrid reactor for pure energy production

    International Nuclear Information System (INIS)

    The next two decades are very critical for nuclear energy development. The commercial fast reactor may be in use around 2035; it is also possible that magnetically confined fusion, laser fusion and z-pinch fusion will be demonstrated at that time. A fusion demonstration reactor can be a pure fusion or a fusion-fission hybrid. The latter can lower the fusion power and mitigate the radiation damage of high energy neutrons to materials. On the other hand, the supply of deuterium and tritium as fuel for fusion can only last a few hundred years. We describe here a hybrid for pure energy use which can make full use of uranium and is proliferation resistant, as no separation of uranium and plutonium is needed in post-processing. The union of fission, fusion, and a pure energy hybrid can contribute to the large scale use of nuclear energy in the near future, and supply mankind for more than a thousand years. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Wagemans Jan

    2016-01-01

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

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

    OpenAIRE

    Salahuddin Asif; Iqbal Masood

    2013-01-01

    Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor....

  19. Nuclear decay by emission of charged particle-superasymmetric fission process

    International Nuclear Information System (INIS)

    The macro-microscopic method, adapted for superasymmetric fission was applied to the alpha decay and other kinds of charged particles emission which are possible due to the nuclear shell structure. Three macroscopic models (the liquid drop model, the finite range of nuclear forces model and the Yukawa exponential model) are extended for nuclear systems with different charge densities. Various numerical methods for the computation of Coulomb and surface energy of a general shape nucleus are presented along with analytical results for some particular shapes. A phenomenological correction was used to obtain the experimental Q-value. This formalism was applied to the alpha decay from the ground state and from a fission isomeric state. A time dependent Hartree-Fock method is used to estimate the zero vibration energy. A new semiempirical formula giving the best estimates for the alpha decay lifetimes was derived and used to predict new alpha emitters. For this new mode of decay intermediate between alpha decay and the traditional fission, larger probabilities are obtained for the combinations of parent-nucleus-heavy cluster leading to a magic daughter nuclei or not too far from it

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Pathological Findings in Mice Exposed to Fission Neutrons in the Reactor GLEEP

    International Nuclear Information System (INIS)

    Mice have been irradiated with fission neutrons liberated from a converter plate exposed to thermal neutrons generated in the graphite-moderated uranium reactor GLEEP. The exposure was continuous and either for the duration of life or for limited periods of a number of weeks or months. The information on life shortening has largely been published already. New data are presented on pathological findings and causes of death. (author)

  3. Nuclear performance of molten salt fusion--fission symbiotic systems for catalyzed DD and DT reactors

    International Nuclear Information System (INIS)

    The nuclear performance of a fusion-fission hybrid reactor having a molten salt composed of Na-Th-F-Be as the blanket fertile material and operating with a catalyzed DD plasma is compared to a similar system utilizing a Li-Th-F-Be salt and operating with a DT plasma. The production of fissile fuel via the 232Th-233U fuel cycle was considered on the basis of its potential nonproliferation aspects. The calculations were performed using one-dimensional discrete ordinates methods to compare neutron balances, fuel producion rates, energy deposition rates, and the radiation damage in the reactor structure

  4. Use of fission reactors for the generation of electricity in the UK civil programme

    International Nuclear Information System (INIS)

    A statement is made for the information of the public on some basic aspects of nuclear power. The needs of society in the near future are considered, in relation to actual or possible risks inherent in various courses of action. The implications of a programme of energy conservation are discussed. Aspects of the use of possible fuels, or the development of alternative energy sources are examined and compared with the use of nuclear power from fission reactors. The possible hazards from reactor accidents, from radioactive wastes and from the action of terrorist groups, are analysed. (U.K.)

  5. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m-2. Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN2, SiC2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  6. Determination of Nuclear Charge Distributions of Fission Fragments from ^{235} U (n_th , f) with Calorimetric Low Temperature Detectors

    Science.gov (United States)

    Grabitz, P.; Andrianov, V.; Bishop, S.; Blanc, A.; Dubey, S.; Echler, A.; Egelhof, P.; Faust, H.; Gönnenwein, F.; Gomez-Guzman, J. M.; Köster, U.; Kraft-Bermuth, S.; Mutterer, M.; Scholz, P.; Stolte, S.

    2016-03-01

    Calorimetric low temperature detectors (CLTD's) for heavy-ion detection have been combined with the LOHENGRIN recoil separator at the ILL Grenoble for the determination of nuclear charge distributions of fission fragments produced by thermal neutron-induced fission of ^{235} U. The LOHENGRIN spectrometer separates fission fragments according to their mass-to-ionic-charge ratio and their kinetic energy, but has no selectivity with respect to nuclear charges Z. For the separation of the nuclear charges, one can exploit the nuclear charge-dependent energy loss of the fragments passing through an energy degrader foil (absorber method). This separation requires detector systems with high energy resolution and negligible pulse height defect, as well as degrader foils which are optimized with respect to thickness, homogeneity, and energy loss straggling. In the present, contribution results of test measurements at the Maier Leibnitz tandem accelerator facility in Munich with ^{109} Ag and ^{127} I beams with the aim to determine the most suitable degrader material, as well as measurements at the Institut Laue-Langevin will be presented. These include a systematic study of the quality of Z-separation of fission fragments in the mass range 82≤ A ≤ 132 and a systematic measurement of ^{92} Rb fission yields, as well as investigations of fission yields toward the symmetry region.

  7. Determination of Nuclear Charge Distributions of Fission Fragments from ^{235}U (n_th, f) with Calorimetric Low Temperature Detectors

    Science.gov (United States)

    Grabitz, P.; Andrianov, V.; Bishop, S.; Blanc, A.; Dubey, S.; Echler, A.; Egelhof, P.; Faust, H.; Gönnenwein, F.; Gomez-Guzman, J. M.; Köster, U.; Kraft-Bermuth, S.; Mutterer, M.; Scholz, P.; Stolte, S.

    2016-08-01

    Calorimetric low temperature detectors (CLTD's) for heavy-ion detection have been combined with the LOHENGRIN recoil separator at the ILL Grenoble for the determination of nuclear charge distributions of fission fragments produced by thermal neutron-induced fission of ^{235}U. The LOHENGRIN spectrometer separates fission fragments according to their mass-to-ionic-charge ratio and their kinetic energy, but has no selectivity with respect to nuclear charges Z. For the separation of the nuclear charges, one can exploit the nuclear charge-dependent energy loss of the fragments passing through an energy degrader foil (absorber method). This separation requires detector systems with high energy resolution and negligible pulse height defect, as well as degrader foils which are optimized with respect to thickness, homogeneity, and energy loss straggling. In the present, contribution results of test measurements at the Maier Leibnitz tandem accelerator facility in Munich with ^{109}Ag and ^{127}I beams with the aim to determine the most suitable degrader material, as well as measurements at the Institut Laue-Langevin will be presented. These include a systematic study of the quality of Z-separation of fission fragments in the mass range 82le A le 132 and a systematic measurement of ^{92}Rb fission yields, as well as investigations of fission yields toward the symmetry region.

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

    International Nuclear Information System (INIS)

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

  9. Coulomb fission of a charged dust cloud in an afterglow plasma

    Science.gov (United States)

    Merlino, Robert; Meyer, John

    2015-11-01

    A dust cloud of 1 micron diameter silica microspheres was confined in a DC glow discharge dusty plasma in argon at a pressure of 100 mTorr (13 Pa). Laser sheet illumination and a fast video camera (2000 frames/s) was used to record the dynamics of this cloud following the switch-off of the plasma and confining forces. Due to the rapid decay of the plasma, and the substantial residual charge on the particles in the plasma afterglow, the cloud evolved under the mutual Coulomb repulsion forces. A variety of dynamic evolutions were observed with different clouds and under different conditions including, Coulomb explosion and expansion. In one case, the cloud underwent a Coulomb fission process, fragmenting into two clouds. Observations and analysis of this Coulomb fission event will be presented. Work supported by DOE.

  10. Neutronic analysis for the fission Mo-99 production by irradiation of a LEU target at RECH-1 reactor

    International Nuclear Information System (INIS)

    For the purpose of developing the capability to produce fission 99Mo, the Chilean Nuclear Energy Commission is participating in the IAEA Coordinated Research Project: 'Developing Techniques for Small Scale Indigenous Mo-99 Production using LEU Fission or Neutron Activation'. Fission 99Mo will be produced irradiating, at RECH-1 reactor, a target made of a LEU metallic uranium foil held between two concentric aluminum tubes. KAERI will provide the LEU foil. Neutronic calculations were performed to estimate the fission products activity for a 13 grams LEU foil annular target, which will be irradiated at the level power of 5 MW during 48 hours. (author)

  11. Fission of Multiply Charged Cesium and Potassium Clusters in Helium Droplets - Approaching the Rayleigh Limit

    CERN Document Server

    Renzler, Michael; Daxner, Matthias; Kranabetter, Lorenz; Kuhn, Martin; Scheier, Paul; Echt, Olof

    2016-01-01

    Electron ionization of helium droplets doped with cesium or potassium results in doubly and, for cesium, triply charged cluster ions. The smallest observable doubly charged clusters are $Cs_{9}^{2+}$ and $K_{11}^{2+}$; they are a factor two smaller than reported previously. The size of potassium dications approaches the Rayleigh limit nRay for which the fission barrier is calculated to vanish, i.e. their fissilities are close to 1. Cesium dications are even smaller than nRay, implying that their fissilities have been significantly overestimated. Triply charged cesium clusters as small as $Cs_{19}^{3+}$ are observed; they are a factor 2.6 smaller than previously reported. Mechanisms that may be responsible for enhanced formation of clusters with high fissilities are discussed.

  12. Fission of multiply charged alkali clusters in helium droplets - approaching the Rayleigh limit.

    Science.gov (United States)

    Renzler, Michael; Harnisch, Martina; Daxner, Matthias; Kranabetter, Lorenz; Kuhn, Martin; Scheier, Paul; Echt, Olof

    2016-04-21

    Electron ionization of helium droplets doped with sodium, potassium or cesium results in doubly and, for cesium, triply charged cluster ions. The smallest observable doubly charged clusters are Na9(2+), K11(2+), and Cs9(2+); they are a factor two to three smaller than reported previously. The size of sodium and potassium dications approaches the Rayleigh limit nRay for which the fission barrier is calculated to vanish, i.e. their fissilities are close to 1. Cesium dications are even smaller than nRay, implying that their fissilities have been significantly overestimated. Triply charged cesium clusters as small as Cs19(3+) are observed; they are a factor 2.6 smaller than previously reported. Mechanisms that may be responsible for enhanced formation of clusters with high fissilities are discussed. PMID:27035406

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

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

  15. Man-machine communication in reactor control using AI methods

    International Nuclear Information System (INIS)

    In the last years the interest in process control has expecially focused on problems of man-machine communication. It depends on its great importance to process performance and user acceptance. Advanced computerized operator aids, e.g. in nuclear power plants, are as well as their man-machine interface. In the Central Institute for Nuclear Research in Rossendorf a computerized operator support system for nuclear power plants is designed, which is involved in a decentralized process automation system. A similar but simpler system, the Hierarchical Informational System (HIS) at the Rossendorf Research Reactor, works with a computer controlled man-machine interface, based on menu. In the special case of the disturbance analysis program SAAP-2, which is included in the HIS, the limits of menu techniques are obviously. Therefore it seems to be necessary and with extended hard- and software possible to realize an user controlled natural language interface using Artificial Intelligence (AI) methods. The draft of such a system is described. It should be able to learn during a teaching phase all phrases and their meanings. The system will work on the basis of a self-organizing, associative data structure. It is used to recognize a great amount of words which are used in language analysis. Error recognition and, if possible, correction is done by means of a distance function in the word set. Language analysis should be carried out with a simplified word class controlled functional analysis. With this interface it is supposed to get experience in intelligent man-machine communication to enhance operational safety in future. (author)

  16. Fission product transport in the high temperature gas-cooled reactor: Theory, program development and verification by recalculation of experiments

    International Nuclear Information System (INIS)

    The high temperature gascooled reactor (HTGR) reaches a special standard in safety because of its high temperature resistent fuel element. After all the possibility of fission product releases can not be excluded without further investigations for HTGRs. The mechanisms of fission product releases, which occur in case of such hypothetical events, are the subject of this work. The main focus of the investigation is how the fission products, which have been released, are re-adsorpted and prevented through this mechanism from being released in the environment. A strong effect of re-adsorption is expected, because experiments have shown that graphite, which is 100% of the core material, has an excellent capability to hold back fission products. With the program tools developed to calculate the fission product transport mechanisms, the corresponding experiments are recalculated and also fission product release calculations are carried out. (orig./HP)

  17. Fault Diagnosis of Batch Reactor Using Machine Learning Methods

    Directory of Open Access Journals (Sweden)

    Sujatha Subramanian

    2014-01-01

    Full Text Available Fault diagnosis of a batch reactor gives the early detection of fault and minimizes the risk of thermal runaway. It provides superior performance and helps to improve safety and consistency. It has become more vital in this technical era. In this paper, support vector machine (SVM is used to estimate the heat release (Qr of the batch reactor both normal and faulty conditions. The signature of the residual, which is obtained from the difference between nominal and estimated faulty Qr values, characterizes the different natures of faults occurring in the batch reactor. Appropriate statistical and geometric features are extracted from the residual signature and the total numbers of features are reduced using SVM attribute selection filter and principle component analysis (PCA techniques. artificial neural network (ANN classifiers like multilayer perceptron (MLP, radial basis function (RBF, and Bayes net are used to classify the different types of faults from the reduced features. It is observed from the result of the comparative study that the proposed method for fault diagnosis with limited number of features extracted from only one estimated parameter (Qr shows that it is more efficient and fast for diagnosing the typical faults.

  18. Direct physical measurements of independent fission yields at a 1-MW research reactor

    International Nuclear Information System (INIS)

    Over the past 20 yr, the number of nuclear reactors on university campuses in the United States has decreased from >70 to <40. Contrary to this trend, the University of Texas at Austin recently completed construction of a new reactor facility at a cost of $5.8 million. The TRIGA Mark II reactor in this facility will be licensed for 1.1-MW steady-state operation and $3.00 power-pulse transients. The new reactor facility was established to enhance the instructional and research opportunities in nuclear science and engineering for both undergraduate and graduate students at the University of Texas. In addition to neutron activation analysis, programs are being planned and equipment is being designed for neutron depth profiling, prompt gamma activation analysis, neutron radiography, and cold neutron research. Because of continued interest in fission-yield system developed by the author when he was at the University of Illinois. The operation of this unique system for the direct physical measurement of independent yields in thermal-neutron fission is reviewed in this paper

  19. Search for neutrino oscillations at a fission reactor

    International Nuclear Information System (INIS)

    In the Gosgen oscillation experiment flux and energy spectrum of these electron antineutrinos were monitored at two distances from the reactor core (37.9 m, 45.9 m). The detector system, well shielded against cosmic radiation, is based on the detection reaction upsilonsub(e)sup(c) + p->n + esup(*) and an alternating array of liquid scintillators and He wire chambers serves as positron and neutron detectors. As signature for a good event a time and position correlation of the detected neutron and positron is required. In addition the measured position energy spectrum directly reflects the incident neutrino energies. To analyze the data in terms of oscillations, either the spectra measured in each position are compared to theoretical predictions for different oscillation parameters or a relative comparison of both measurements is performed. Allowing in the analysis a variation of 2 standard deviations for the systematic uncertainties the following limits on the oscillation parameters are obtained in a combination of the data from both measuring positions: mixing angle: [large mass param.] sin2 20 2 2 (90% c.l.). (orig./HSI)

  20. Corrosion behaviour of the tool steel of the fuel charge machine during cleaning process

    International Nuclear Information System (INIS)

    In the framework of the experimentation activity of the PEC Reactor components, presently in course at the Casaccia Energy Research Centre (CRE Casaccia), the sodium removal process has been usually carried out by means of Butylcellosolve. Recently, in order to eliminate flammable organic solvent from the reactor building, it was decided to evaluate the possibility of using an atomized water method for the cleaning of the Fuel Charge Machine (FCM). Two important problems have been immediately identified: - lower removal process efficiency; since the geometries of the PEC Reactor FCM are complex, there are a number of areas of retention, where liquid access is difficult, - component damage due to the corrosion process. The main risk is associated with the formation of aqueous NaOH which can give rise to caustic stress corrosion cracking. In order to know something more about the above mentioned problems, a test programme was designed whose main aims were: - cleaning tests efficiency determination of gripper prototype by atomized water method using nitrogen gas or alternatively carbon dioxide; - study of the corrosion behaviour of tool steel in caustic solutions. This paper reports the results of the corrosion tests

  1. Application of STAV5 code for the analysis of fission gas release in power reactor rods

    International Nuclear Information System (INIS)

    STAV5 is a design code for calculation of temperatures, fission gas release and rod pressure in BWR and PWR fuel rods. It includes the treatment of pellet cracks affecting conductivity and thermal expansion, gap closure by eccentric or relocated pellet fragments and oxide and crud build-up on the clad outer surface. The fission gas release model consists of two parts: High temperature release based on grain boundary saturation and low temperature release varying with fission rate of different isotopes. STAV5 has been benchmarked with a number of inpile thermal measurement experiments to as high burnup as 25 MWd/kg U. The main application of STAV5 is as a routine design tool for power reactor rods. It is also used to compare with PIE data. Examples are given from the analyses of fission gas release data from BWR rods from Oskarshamn 1 and Barsebeck 1 as well as PWR rods from Maine Yankee initial cores. The STAV5 evaluation show the importance of power histories, densification and the position in the assembly. (author)

  2. A Multichannel Analysis System for Low Levels of Fission Products in Reactor Coolants

    International Nuclear Information System (INIS)

    A multi-channel analysis system consisting of a 5-in x 4-in. NaI(Tl) crystal and a 256-channel pulse height analyser incorporating ferrite core memory storage is described. This system has been adopted for the measurement of very low levels of fission products in reactor coolants by gamma spectrometry. The pulse height analyser is based on a Hutchinson and Scarrot type analog-to-digital converter using a 2-Mc clock. Binary address and storage system are used for the computer part of the analyser. The memory has a sixteen bit storage capacity. The memory cycle time is 20 μs. Two analog output modes, i.e. a CRT and an X-Y recorder, are provided for rapid identification of the isotopes from the gamma spectra. For accurate measurement and analysis of activities two digital output modes are provided, namely a digital printer and a neon read-out. The system can be used for a very large range of fission-product activity, from very low levels of activity encountered in normal operations to fairly high levels encountered in the case of a split rod. Using the above system, preliminary studies have been made on induced radioactivity and fission-product activities in the reactor cooling water samples. (author)

  3. Characteristics of fission products behavior on a severe accident in fast breeder reactor

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been developing the ACTOR code for the analysis of the fission products behavior under the severe accident condition to apply the probabilistic safety assessment to fast breeder reactor plants. Major analysis models of the ACTOR code were validated and adjusted by related experimental results. The fission products behavior on PLOHS (Protected Loss of Heat Sink) sequence which is one of the typical severe accidents in FBR plant was analyzed by using the ACTOR code. It was confirmed that the ACTOR had an enough capability to analyze the fission products behavior during severe accident. From the analysis results of PLOHS, it was confirmed that cesium is transferred to the cover gas region much more than iodine because iodine which is one of halogen connects to sodium easily and is retained in the coolant. Therefore, cesium is important and it is needed to examine the necessity to treat cesium as one of FPs considered in reactor establishment permission for FBR plant. Thus, cesium transfer behavior in sodium during the rare gas bubbles rise from fuel to the cover gas region was confirmed to be very important. And JNES started study including validation test about cesium transfer behavior with Hokkaido University. (author)

  4. ACRR [Annular Core Research Reactor] fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs

  5. Partitioning and transmutation of 99Tc in fission reactors and hybrids systems

    International Nuclear Information System (INIS)

    Partitioning and transmutation of radioactive and long lived component from the highly radioactive waste stream in order to reduce or probably eliminate their radio toxic inventory was the important option for the nuclear waste management. The important fission products that deserve most attention is the technetium. Technetium is present as a single isotopic species (99Tc) can be transmuted by single neutron capture into the stable noble metal ruthenium ( 100Ru). The technetium separation from spent fuel is possible with PUREX process. An, other chemical process was developed to separate a priory technetium with uranium is the UREX process. The transmutation of 99Tc in thermal reactor such as LWRs will be difficult because of the long transmutation half-lives and the large inventory required. Better result can be obtained in fast reactors, or in accelerator driven height flux reactor

  6. Non linear seismic analysis of charge/discharge machine

    International Nuclear Information System (INIS)

    The main conclusions of the seismic analysis of the Latina CDM are: i. The charge machine has been demonstrated to be capable of withstanding the effects of a 0.1 g earthquake. Stresses and displacements were all within allowable limits and the stability criteria were fully satisfied for all positions of the cross-travel bogie on the gantry. ii. Movements due to loss of friction between the cross-travel bogie wheels and the rail was found to be small, i.e. less than 2 mm for all cases considered. The modes of rocking of the fixed and hinged legs preclude any possibility of excessive movement between the long travel bogie wheels and the rail. iii. The non-linear analysis incorporating contact and friction has given more realistic results than any of the linear verification analyses. The method of analysis indicates that even the larger structures can be efficiently solved on a mini computer for a long forcing input (16 s). (orig.)

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

    International Nuclear Information System (INIS)

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

  8. The effect of intermittent operation on local fission rate in the McMaster Nuclear Reactor

    International Nuclear Information System (INIS)

    The McMaster Nuclear Reactor operates on a 16-hr/day/5-day/week schedule causing cyclic loading of Xenon in the core and requiring compensation by the control systems to maintain operations. The constant control rod interaction affects local fission rates. This paper confirms the relationship between Xenon load and control rod movement and studies the relationship between local fission rate and control rod insertion. The results provide information related to analysis approximations used in depletion calculations. In addition, comparisons are made between the current MNR operation cycle and a proposed continuous operational approach. The results are further discussed in the context of proposed Molybdenum-99 production at MNR. (author)

  9. Integral measurement of fission-product reactivity worths in some fast reactor spectra

    International Nuclear Information System (INIS)

    The reactivity worth per atom for a number of fission-product isotopes relative to that of 235U was measured in three various fast-reactor spectra. The following isotopes were studied: 95Mo, 97Mo, 99Tc, 101Ru, 102Ru, 104Ru, 103Rh, 133Cs, 147Pm and 149Sm. A fission product mock-up sample was also included in the measurements. The reactivity worths were measured by the pile-oscillator technique. The fundamental mode amplitude of the perturbation signal was obtained through Fourier analysis. The experimental results are compared with calculated values obtained from perturbation calculations using published cross-sections for the sample materials. From a comparison between the measured and the calculated reactivity worths it is concluded that only the 95Mo, 104Ru and 149Sm worths are well predicted in all three systems. For the other samples, the calculated values are generally too high. (author)

  10. The man - machine - organization system analysis for research reactor

    International Nuclear Information System (INIS)

    The man-machine - organization (MMO) system analysis integrates into the human factor analysis methodology for the complex installation operation. In order to perform such analysis it is necessary the interfaces analysis from MMO system by using the Human Reliability Analysis (HRA) methods. The purpose of this paper is to identify man-machine - organization interfaces that could lead to an accident, types of human interactions which may mitigate or exacerbate the accident, types of human errors, performance shaping factors and to estimate of human error probability as effects of human performance in reliability and safety. The results of this paper are the interfaces that could have a major contribution to the human error probabilities. Conclusively, some modifications are recommended in MMO system in order to reduce the human error probabilities and the contribution of the human factor to system unavailability. The necessary information was obtained from operating experience of research reactor TRIGA from INR Pitesti. The required data were obtained from generic data bases. (authors)

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

    International Nuclear Information System (INIS)

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

  12. EPRI Asilomar papers: on the possibility of advanced fuel fusion reactors, fusion-fission hybrid breeders, small fusion power reactors, Asilomar, California, December 15--17, 1976

    International Nuclear Information System (INIS)

    An EPRI Ad Hoc Panel met in Asilomar, California for a three day general discussion of topics of particular interest to utility representatives. The three main topics considered were: (1) the possibility of advanced fuel fusion reactors, (2) fusion-fission hybrid breeders, and (3) small fusion power reactors. The report describes the ideas that evolved on these three topics. An example of a ''neutron less'' fusion reactor using the p-11B fuel cycle is described along with the critical questions that need to be addressed. The importance to the utility industry of using fusion neutrons to breed fission fuel for LWRs is outlined and directions for future EPRI research on fusion-fission systems are recommended. The desirability of small fusion power reactors to enable the early commercialization of fusion and for satisfying users' needs is discussed. Areas for possible EPRI research to help achieve this goal are presented

  13. Optimal design study of cylindrical finned reactor for solar adsorption cooling machine

    Energy Technology Data Exchange (ETDEWEB)

    Allouache, N. [Univ. des Sciences et de la Technologie Houari Boumediene, Bab Ezzouar (Algeria). Faculte de Genie Mecanique et de Genie des Procedes; Al Mers, A. [Moulay Ismail Univ., Meknes (Morocco). Ecole National Superieure d' Art et Metiers

    2010-07-01

    Solid adsorption cooling machines use medium temperature industrial waste heat together with a renewable energy source, such as solar energy. The adsorption cooling machine consists of an evaporator, a condenser and a reactor containing a solid adsorbent. In this study, a model was developed for thermodynamic performance analysis and optimization of a cylindrical finned solar reactor in an adsorption refrigerator working with activated carbon-ammonia. The heat and mass transfer in the adsorption cooling machine was determined. The model was validated using experimental results. The study investigated the sensitivity of the machine performance versus the geometrical configuration of the reactor. The study showed that for an optimized reactor, a higher fin number significantly reduces the heat losses of the reactor. It was concluded that the solar coefficient of performance (COP) of an optimized reactor can reach 45 per cent when the number of fins varies between 5 and 6. 10 refs., 4 figs.

  14. Technical bases to consider for performance and demonstration testing of space fission reactors

    International Nuclear Information System (INIS)

    Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as: (1) Do the benefits outweigh the risks; (2) Are there equivalent alternatives; (3) Can a test facility be constructed (or modified) in a reasonable amount of time; (4) Will the test article accurately represent the flight system; and (5) Are the costs too restrictive, have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems.

  15. Fission product release from a pressurized water reactor defective fuel rod: effect of thermal cycling

    International Nuclear Information System (INIS)

    The emission of fission gases and iodines by a pressurized water reactor fuel rod containing a defect when it is initially put in the reactor is studied experimentally using a pressurized water loop in the Siloe reactor at Grenoble. The initial leakage is simulated by making a small hole near the upper end of the rod. The rare gases and iodines are continuously analyzed, and the source terms of fission products are expressed as the ratio of the release rate of a given isotope from the defective fuel rod to the birth rate of this isotope. The release fractions of rare gases and iodines have been determined in different conditions: steady power level between 120 and 700 W . cm-1, power cycling in the range of 200 to 400 W . cm-1, and in the range 120 to 400 W . cm-1. At steady power level, the amounts of radioactive gases escaped from the rod are 100 times higher than those emitted by a sound fuel submitted to a similar power level. The power cycling favors the emission of all iodines whose release rate is 10 to 20 times higher than at the maximum steady power level

  16. Lasers from fission (gaseous core reactors and nuclear pumped lasers for space power generation and transmission)

    International Nuclear Information System (INIS)

    The energization of lasers directly by nuclear reactions has recently been achieved. In experiments conducted jointly by the University of Florida and Los Alamos Scientific Laboratory, New Mexico, a helium-xenon laser was directly pumped by fission fragments. The obtained laser wavelength was 3.5 μm. A group of researchers at the Sandia Corporation in Albuquerque, New Mexico, was successful in energizing a carbon monoxide laser by fission fragments at wavelengths in the 5-μm band. At the University of Illinois lasing was achieved at wavelengths of 8629 A and 9393 A in a neon-nitrogen mixture. A program of gaseous core reactor research is underway with experiments being conducted at the Los Alamos Scientific Laboratory in New Mexico, U.S.A. The program utilizes a beryllium moderator-reflector, forming a cylindrical cavity of 1 m diameter and 1 m length. This system and associated control system hardware, uses components from the previous ROVER nuclear rocket program. Various configurations of canisters containing enriched gaseous uraniumhexafluoride fuel are inserted into the reactor cavity for research on neutronics and nuclear induced optical radiation. Critical mass, control swing and the effects of poison were measured by simulating enriched uranium hexafluoride fuel with uranium foils, which were placed in homogeneous and inhomogeneous distributions in the cavity. Critical mass was determined at about 6 kg 93% enriched 235 uranium. A uranium hexafluoride canister system was built for safe operation in the reactor cavity and for physics measurements and observations at nuclear criticality. It is anticipated that this work will result in the demonstration of principles of a new type of nuclear power reactor, and of laser output from such a reactor. (author)

  17. Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

    International Nuclear Information System (INIS)

    Prior work has demonstrated the use of a natural B4C capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method to critical assemblies for performing nuclear data measurements at near 235U fission-energy neutron spectrum. Previous fission product measurements showed that the neutron spectrum achievable with natural B4C was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a B4C capsule 96 % enriched in 10B resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. Fission product yields measured following an irradiation of a sample with this new method and subsequent radiochemical separations are presented here. (author)

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

    International Nuclear Information System (INIS)

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

  19. Method for supervising and controlling the charging and discharging operations of fuel in a nuclear reactor and apparatus for applying this method

    International Nuclear Information System (INIS)

    Method is described for supervising and controlling the charging and discharging operations, in nuclear fuel element assemblies, of a core of a reactor, of a reactor pond and of a decontamination pond, by means of a charging machine equipped with a telescopic mast the end of which is provided with a gripping head with grippers, serving the reactor and the reactor pond in which there is arranged a buffer storage rack, a fixed depositing station and a mobile depositing station, and by means of a charging machine equipped with a telescopic mast the end of which is provided with a gripping head with grippers, serving the decontamination pond containing storage racks, and by means of a transfer device providing communication between the reactor pond and the decontamination pond, characterised in that the initial position in each assembly in the core of the reactor, in the storage racks and possibly in the buffer rack is recorded, in that the position of the charging machine and/or of the handling machine and/or of the transfer device and/or of the mobile depositing station is recorded, likewise the identification of the assembly at the time of each taking up of an assembly and/or at the time of each placing of an assembly in the core of the reactor, in the buffer rack in the transfer basket, in the storage rack, in the fixed depositing station and in the mobile depositing station, in that the command and control signals for each manipulation required of the charging machine, of the handling machine, of the transfer device and of any other mobile station are compared with the recorded signals of a preestablished charging sequence. 5 refs., 4 figs

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

    International Nuclear Information System (INIS)

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S2CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

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

    International Nuclear Information System (INIS)

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

  2. New experimental set-ups for investigating properties of the spontaneously fissionable 242Am isomer in the IBR-30 and IBR-2 pulsed reactors

    International Nuclear Information System (INIS)

    The mean neutron number per isomeric fission event of 242Am is measured. Correlations between the isomeric fission probabilities and the fission widths of neutron resonance of 241Am are investigated. For these experiments neutron time-of-flight spectrometers based on pulsed reactors IBR-30 and IBR-2 at the JINR Dubna are used

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

    International Nuclear Information System (INIS)

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

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

    CERN Document Server

    Calviani, Marco; Montagnoli, G; Mastinu, P

    2009-01-01

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

  5. Tests on a prototype of the Passive Fission Gas Monitor for failed detection (PRISM reactor)

    International Nuclear Information System (INIS)

    The Passive Diffusion Fission Gas Monitor PDFGM is mounted on the PRISM reactor head and extends into the cover gas Region where it determines the presence of radioactive fission gases (Kr, Xe, and so on) released from failed fuel pins. It contains a steel diffusion column that is closed at the upper end but opened to the cover gas at its lower end. The upper portion of the column is located in the field of view of a collimated gamma detector which is shielded from the remainder of the cover gas and of the sodium pool below. Heaters are provided to obtain a uniform axial temperature in the gas column and to minimize the potential for natural convection currents. In this way, the molecular diffusion can be established based on the fission gas concentration gradients along the column length. This is an advanced solution in comparison with current devices based on active components (pumps, filters, and so on). The experimental results on a prototype of PDFGM and their interpretation will be presented in this paper. (author)

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

    International Nuclear Information System (INIS)

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

  7. The different facilities of the reactor PHENIX for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

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

  8. Multidimensional analysis of fission gas discharge following fuel element failure in sodium fast reactor

    International Nuclear Information System (INIS)

    The required technological and safety standards for future Gen IV. Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The development of a suite of high performance computational tools for multiscale simulations of Gen-IV Sodium Fast Reactor (SFR) has recently been undertaken by a DOE-sponsored university consortium . The purpose of this paper is to present the modeling concept for mechanistic computer simulations of the injection of a jet of gaseous fission products into a partially blocked SFR coolant channel following localized cladding overheat and breach. A three-dimensional model of gas/liquid-sodium interaction has been developed based on a multifield modeling framework implemented in the NPHASE-CMFD code. The boundary conditions used as input to NPHASE-CMFD have been obtained by averaging the results of DNS simulations performed using the PHASTA code. (author)

  9. Fission product and chemical energy releases during core melt events in U-Al research reactors

    International Nuclear Information System (INIS)

    Fission product releases data from heated uranium-aluminum reactor fuels are analyzed. Extensive library of correlations was developed for predicting releases which may vary with time, burnup ambient, fuel-type subject to certain assumptions. Correlations were developed in various forms for U-Al. (dispersed/alloy), U308-Al (dispersed) and dispersed U3Si2-AL, and U3Si-Al Fuels. Overall statistics is quite favorable. Unresolved issues and data needs demand best estimate analyses of reactors using U3Si2-Al fuel. Importance of capturing fragment size distribution was demonstrated. Results agree with Nelson's observations for onset ignition. The need to develop an appropriate fragmentation model was evident

  10. Fission-product aerosol sampling system for LWR experiments in the TREAT reactor

    International Nuclear Information System (INIS)

    This work summarizes the design and collection characteristics of a fission-product aerosol sampling system that was developed for a series of light water reactor (LWR) source-term experiments under consideration for performance in 1984 at Argonne National Laboratory's TREAT reactor. These tests would be performed using a bundle of four preirradiated, Zircaloy-clad LWR fuel pins. In these tests, fuel pin integrity would be breached under various simulated accident conditions. The aerosol sampling system was designed to efficiently extract and collect these aerosols such that time-averaged aerosol size distributions, number concentrations and mass loadings could be determined accurately for each experiment, using a combination of real-time and time-interval measurements and post-test analytical techniques. The entire system also was designed to be disassembled remotely because of potentially high levels of radioactivity

  11. Role of Fission Reactors and IFMIF in the Fusion Materials Programme

    International Nuclear Information System (INIS)

    In fusion power reactors, the plasma facing (first wall and divertor) and breeding blanket components will suffer irradiation by an intense flux of 14.1 MeV neutrons coming from the plasma. These fusion neutrons will produce nuclear transmutation reactions and atomic displacement cascades causing the presence of impurities and defects. Therefore, the chemical composition and the microstructure of the materials will change after irradiation, affecting its physical and mechanical properties. The study and evaluation of the changes in the material properties under irradiation is a top priority for the design of a fusion reactor. Key irradiation parameters include the accumulated damage, expressed in the number of displacements per atom or dpa, the damage rate in dpa/s, the rates of production of impurities (e.g. ppm(He)/dpa and ppm(H)/dpa ratios) and the temperature of the materials under irradiation. Unfortunately, at the moment, the existing sources of 14 MeV neutrons have very small intensity and do not allow us to get significant damage accumulation in a reasonable time. Therefore, it is necessary to simulate irradiation by fusion neutrons through the use of fission neutrons, high energy protons or heavy ions. Although the irradiation conditions provided by such particles are very different from those expected to occur in a fusion power reactor, especially in terms of damage rate and rates of production of impurities, relevant information can be obtained from present available fission reactors. In the paper a list with relevant experiments suitable for the fusion community is given, and the role of the future International Fusion Materials Irradiation Facility is discussed. (author)

  12. Neutron Damage in the Plasma Chamber First Wall of the GCFTR-2 Fusion-Fission Hybrid Reactor

    Science.gov (United States)

    Pinto, L. N.; Gonnelli, E.; Rossi, P. C. R.; Carluccio, T.; dos Santos, A.

    2015-07-01

    The successful development of energy-conversion machines based on either nuclear fission or fusion is completely dependent on the behaviour of the engineering materials used to construct the fuel containment and primary heat extraction systems. Such materials must be designed in order to maintain their structural integrity and dimensional stability in an environment involving high temperatures and heat fluxes, corrosive media, high stresses and intense neutron fluxes. However, despite the various others damage issues, such as the effects of plasma radiation and particle flux, the neutron flux is sufficiently energetic to displace atoms from their crystalline lattice sites. It is clear that the understanding of the neutron damage is essential for the development and safe operation of nuclear systems. Considering this context, the work presents a study of neutron damage in the Gas Cooled Fast Transmutation Reactor (GCFTR-2) driven by a Tokamak D-T fusion neutron source of 14.03 MeV. The theoretical analysis was performed by MCNP-5 and the ENDF/B-VII.1 neutron data library. A brief discussion about the determination of the radiation damage is presented, along with an analysis of the total neutron energy deposition in seven points through the material of the plasma source wall (PSW), in which was considered the HT-9 steel. The neutron flux was subdivided into three energy groups and their behaviour through the material was also examined.

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

    International Nuclear Information System (INIS)

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

  14. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    Science.gov (United States)

    Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

    2011-01-01

    The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

  15. Startup of the Fission Converter Epithermal Neutron Irradiation Facility at the MIT Reactor

    International Nuclear Information System (INIS)

    A new epithermal neutron irradiation facility, based on a fission converter assembly placed in the thermal column outside the reactor core, has been put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). This facility was constructed to provide a high-intensity, forward-directed beam for use in neutron capture therapy with an epithermal flux of [approximately equal to]1010 n/cm2.s at the medical room entrance with negligible fast neutron and gamma-ray contamination. The fission converter assembly consists of 10 or 11 MITR fuel elements placed in an aluminum tank and cooled with D2O. Thermal-hydraulic criteria were established based on heat deposition calculations. Various startup tests were performed to verify expected neutronic and thermal-hydraulic behavior. Flow testing showed an almost flat flow distribution across the fuel elements with <5% bypass flow. The total reactivity change caused by operation of the facility was measured at 0.014 ± 0.002% δK/K. Thermal power produced by the facility was measured to be 83.1 ± 4.2 kW. All of these test results satisfied the thermal-hydraulic safety criteria. In addition, radiation shielding design measurements were made that verified design calculations for the neutronic performance

  16. Study on fusion-fission hybrid reactor for transmutation of nuclear waste

    International Nuclear Information System (INIS)

    A conceptual design of fusion-fission hybrid reactor (FFHR) has been studied for transmutation. The FFHR is magnetic confinement fusion reactor with a blanket containing minor actinides (MAs) from nuclear waste reprocessed from three years cooled spent PWR fuels of 33 GWD/t with initial uranium enrichment of 3.3%. The same plasma conditions as ITER were referred for blanket design. The blanket system consists of MA pebbles coated with Ti nitride, cooling water, structure (F82H) and tritium breeding materials (Li2TiO3). A criticality (keff), Tritium Breeding Ratio (TBR) and energy multiplication (Q value) in the blanket were investigated by the Monte Carlo N-particle transport code MCNP-4C with the nuclear data library JENDL 3.3 after understanding heat calculations. The Q value of 105, fission power 1200 MWt and TBR of 1.8 were obtained at the both conditions of keff value of 0.95 and permissible temperatures in MA, F82H structure, Li2TiO3 and coolant. (author)

  17. Safety and economical requirements of conceptual fusion power reactors in co-existing advanced fission plants

    International Nuclear Information System (INIS)

    An EPR fission plant is expected to operate from 2010 to 2070. In this time range a new generation of advanced fission reactors and several stages of fusion reactors from ITER to DEMO will emerge. Their viability in the competitive socio-economic environment and also their possible synergy benefits are discussed in this paper. The studied cases involve the Finnish EPR, Generation IV, and the EFDA Power Plant Conceptual Study Models A-D. The main focus is on economic and safety assessments. Some cross-cutting issues of technologies are discussed. Concerning the economic potential of both conceptual fusion power plants and those of Generation IV candidates, we have used the present Finnish EPR as a reference. Comparisons using various pricing methods are made for fusion and Generation IV: mass flow analyses together with engineering, construction and financial margins form one method and another one is based on simple scaling relations between components or structures with common technology level. In all these studies fusion competitiveness has to be improved in terms of plant availability and internal power recirculation. At present the best fission plants have a plant availability close to 95% and an internal power recirculation of the order of 3-4%. The operation and maintenance solutions of Model C and D show the right way for fusion. A remarkable rise of the fuel costs of present LWRs would first make the Generation IV breeder options and thereafter the fusion plants more competitive. The costs of safety related components, such as the containment and the equipment for severe accident mitigation (e.g. the core catcher in a LWR), should be accounted for and the extent to which the inherent fusion safety features could compensate such expenses should be analysed. For an overall assessment of the various nuclear options both internal and external costs are considered. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    A systematic study about the registration characteristics of synthetic fused quartz 'Suprasil I' use as a detector of fission fragments under high flux of reactor neutrons and the effects of irradiation on it was performed. Fission fragments of 252Cf, gamma radiation doses of of 60Co up to 150 MGy, and integrated neutrons fluxes up to 1020 n/cm2 were used. A model to explain the effects on track registration and development characteristics of 'Suprasil I' irradiated on reactors were proposed, based on the obtained results for efficiency an for annealing. (C.G.C.)

  20. Primary Distributions of Nuclear Charge for Fission-Fragment Masses 132, 134, 136 and 137 from Thermal Fission of U235

    International Nuclear Information System (INIS)

    By a mass spectrometer fission fragments from thermal fission of U235 are exactly separated with respect to mass and kinetic energy within a time of 10-6 s after fission. The separated fragments are caught in a β-sensitive Ilford G 5 emulsion that is located in the focal plane of the spectrometer. Development of the irradiated emulsions is carried out, if possible, after a time long compared with the longest half-life of the regarded decay chain. Half-lives of days or longer are not taken into account, but corrections can be easily made for them. After development of the emulsions all beta tracks emerging from the end of every fission-fragment track can be seen under the microscope. The possibility of correlating every single β-track with a particular fission-fragment track allows the evaluation of the number n(x) of fission fragments possessing x β-tracks, thus giving not only the mean chain length but also the β-particle distribution. As the stable end product of each decay chain is known, this β-distribution is an exact image of the primary nuclear charge distribution. In the measurements done up to now only β-particles emitted into the half solid angle formed by the emulsion plate were registered, buta simple statistical calculation enables the desired 4π-distribution to be evaluated. By this method β-distributions at fixed kinetic energies near the mean kinetic energy of each fragment mass are given for the masses 132, 134, 136 and 137. For the lower masses 132 and 134 the neutron shell N = 82 is responsible for the most probable primary charges near 50 and 52 respectively. For M = 136 and 137 the primary charge is about 53 and 53.2. Additional approximative corrections in respect of conversion electrons (by omitting very short β-tracks corresponding to very low β-energies) and to delayed neutrons (for mass 137) were not very large. Similar measurements carried out directly in 4π—geometry to avoid the statistical error arising from the

  1. Final report on study of advanced fission power reactor development for the United States. Volume III

    International Nuclear Information System (INIS)

    This three-volume set details a multistage research study on the comparison of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. Volume III presents the basic data and other input information utilized in the process of the study. Detailed reactor and fuel-cycle information is contained on the HTGR, GCFR, MSBR, LWBR, and CANDU-HWR, as obtained by BCL staff from numerous sources. These data have been assembled, critically reviewed, and modified where necessary to assure consistency, both internally for each reactor system, and between reactor types. Similar, comparative information was required for the LWR mass balances, fuel cycle, etc., but this was generally accepted from ERDA sources without the critical evaluation to which the other data were subjected. Other detailed information that relates to the objective of the study, e.g., projections of nuclear power growth and uranium availability, generic safeguards information, description of computerized models, etc., is incorporated for ready reference

  2. Improving Nuclear Safety of Fast Reactors by Slowing Down Fission Chain Reaction

    Directory of Open Access Journals (Sweden)

    G. G. Kulikov

    2014-01-01

    Full Text Available Light materials with small atomic mass (light or heavy water, graphite, and so on are usually used as a neutron reflector and moderator. The present paper proposes using a new, heavy element as neutron moderator and reflector, namely, “radiogenic lead” with dominant content of isotope 208Pb. Radiogenic lead is a stable natural lead. This isotope is characterized by extremely low micro cross-section of radiative neutron capture (~0.23 mb for thermal neutrons, which is smaller than graphite and deuterium cross-sections. The reflector-converter for a fast reactor core is the structure capable of transforming some part of prompt neutrons leaked from the core into the reflected neutrons with properties similar to those of delayed neutrons, that is, sufficiently large contribution to reactivity at the level of effective fraction of delayed neutrons and relatively long lifetime, comparable with lifetimes of radionuclides-emitters of delayed neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in the fast reactor. This can improve the fast reactor safety and reduce some requirements to the technologies used to fabricate fuel for the fast reactor.

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

    International Nuclear Information System (INIS)

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

  4. Heterogeneity and alteration of uraninite from the natural fission reactor 10 at Oklo, Gabon

    International Nuclear Information System (INIS)

    A mineralogical study of uranium ore from reactor zone 10 revealed that uraninite in the natural reactors at Oklo, Gabon, has been altered through partial dissolution, Pb loss, and replacement by coffinite, USiO4.nH2O. The dissolution occurred during formation of the clay mantle surrounding the ore body and was probably caused by hydrothermal saline solutions under reducing conditions. The loss of lead (up to 11 wt%) from uraninite occurred approximately one billion years after the operation of the reactors. As a result, there are two generations of uraninite in the reactor zone that differ in chemical composition and unit cell parameters [a1 = 0.5495(2) and a2 = 0.5455(2) nm]. Minor coffinitization of uraninite has also occurred. Thus, the Oklo deposit has been altered since the event of nuclear criticality. This provides several distinct geochemical environments in which one may analyze the corrosion of uraninite and the subsequent retention or migration of fission products. (author). 20 refs., 3 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

  6. Multidimensional analysis of fission gas transport following fuel element failure in sodium fast reactor

    International Nuclear Information System (INIS)

    Highlights: ► High performance computing applied to Gen. IV SFR reactor accident predictions. ► Approach: multiscale combined DNS and RANS models of two-phase flow. ► Simulation of fission product gas injection into partially-blocked coolant channel. ► Physical consistency and numerical accuracy of the method have been demonstrated. - Abstract: Significant progress in several areas will have to be made to achieve the required technological and safety standards for future Gen. IV reactors, including both novel experimental methods (starting with separate-effect, then followed by integral experiments) and high performance computational models characterized by the necessary level of modeling detail and high accuracy of predictions. Furthermore, it is important that the experimental and theoretical/computational research complement each other, so that the results of measurements could be directly used for model validation purposes, whereas the results of simulations should provide input to identify modeling uncertainties and provide guidelines for prioritizing future experiments. The purpose of this paper is to present the modeling concept for mechanistic computer simulations of the injection of a jet of gaseous fission products into a partially blocked SFR coolant channel following localized cladding overheat and breach. A three-dimensional model of gas/liquid-sodium interaction has been developed based on a multifield modeling framework implemented in the NPHASE-CMFD code. The boundary conditions used as input to NPHASE-CMFD have been obtained by averaging the results of direct numerical simulations (DNS) performed using the PHASTA code. The novel aspects of the results discussed in the paper include the demonstration of advantages of using a multiscale approach to model local phenomena governing gas/liquid-sodium two-phase flow inside reactor coolant channels following cladding breach, as well as the observations about areas where future experiments are

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

    International Nuclear Information System (INIS)

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

  8. Development of high temperature fission counter-chamber(FC)S for a top entry loop type fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype high temperature fission counter-chambers have been made as neutron detectors for installation in the reactor vessel of the 600MWe-class top entry loop type fast breeder reactor. Using these prototypes as samples, a high-temperature endurance test has been conducted. The validity of the prototypes has been established by the test results, which show that the prototypes nearly satisfy the design performance. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. Reaction balance and efficiency analysis of a D-D fusion/fission hybrid with satellite D-3He reactors

    International Nuclear Information System (INIS)

    Selected reactor physics and isotope balance characteristics of a fusion hybrid supported D-3He satellite nuclear energy system are formulated and investigated. The system consists of two types of reactors: a parent D-fueled fusion device and a number of smaller reactors optimized for D-3He fusion. The parent hybrid station breeds the helium-3 for the satellites and also breeds fissile fuel for an existing fission reactor economy. Various hybrid operational regimes are examined in order to determine favorable reactor Q values and effective fusion and fission efficiencies. A number of analytical correlations between power output, plasma energetics, blanket neutronics, breeding capacity, and energy conversion cycles are established and evaluated. Numerical examples of performance parameters such as fission-tofusion power, overall conversion efficiency, and the ratio of satellite to parent fusion power are presented. The range of reactor efficiencies is elucidated as affected by the internal plasma power balances. As an upper bound based on optimistic injection and direct conversion efficiencies, we find the D-3He satellite system power output attaining at best 1/3 of the parent fusion power

  13. Fission fragments and structural materials cross sections accuracy influence on main characteristics in fast neutron reactor physics

    International Nuclear Information System (INIS)

    This report is devoted to the exactingness in neutronic data to impose at the fission products and at the structural materials. To construct fast neutron power reactors it is necessary to know the accuracy in the calcul of multiplication factor, regeneration factor and reactivity losses. In this report the main error sources are examined

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

    International Nuclear Information System (INIS)

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

  15. The optimization of the combination of various reactors and the important role of fusion-fission hybrid reactors in the development of nuclear energy in China

    International Nuclear Information System (INIS)

    For energy demand in the economic development of China in 21 Century, for seeking the strategy to develop nuclear energy in China, according to the nuclear resources in China and the perspective of international nuclear technology development, the optimization of the combination of three kinds of advanced reactors, namely, HTGR, FBR, and fusion-fission hybrid reactors in the development of nuclear energy in China was investgated. Three alternative stra tegies with different priorities were suggested

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

    International Nuclear Information System (INIS)

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

  17. Models of fission gas behaviour in fast reactor fuels under steady state and transient conditions

    International Nuclear Information System (INIS)

    Two different physical mechanisms have been used in the past as the basis of models describing fission gas release and swelling in an operating fast reactor fuel. This has led to confusion in the interpretation of experimental data and to a proliferation of apparently conflicting models. This work aims at resolving some of these difficulties and shows clearly that the real situation can only be described completely by a model which combines the two basic mechanisms, each dominating in a different temperature regime. The rate theory is used to describe the nucleation and evolution of the fission-gas bubble population. At lower temperatures the model is based upon the notion of the random motion of single gas atoms and gas bubbles and includes the effects of re-solution and coalescence within the bubble population whilst at higher temperatures the directed motion of bubbles in a temperature gradient, again including re-solution and coalescence, is shown to be the most important mechanism. There are further difficulties within each of these overall schemes and in particular the sensitivity of the results of the random motion model to the re-solution rate and bubble mobility are highlighted. (Auth.)

  18. Fission-product chemistry in severe reactor accidents: review of relevant integral experiments

    International Nuclear Information System (INIS)

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

  19. Ternary Fission

    International Nuclear Information System (INIS)

    The fission process in which heavy nuclei fragment into three large charged panicles, in place of the usual two, has been studied in the case of thermal-neutron-induced fission of U235 and the spontaneous fission of Cf252. Solid-state detectors, a fast triple coincidence system and a three-coincident-parameter analyser were used to measure the three fission fragment energies parallel with the detection of each ternary fission event. Experimental evidence is presented supporting the existence of ternary fission by specifically excluding recoil phenomena and accidental events as contributing to the observed three-fold coincidence events. Mass-energy-angular correlations of ternary fission have been determined and are summarized as follows: The total kinetic energy release in ternary fission appears to be slightly higher (by approximately 10 MeV) than that for binary fission. In the case of the spontaneous ternary fission of Cf252, the frequency of occurrence is observed to be greater than 2.2 x 10-6 ternary fission events per binary fission event. Tripartition of Cf252 results preferentially in division into two medium mass particle (one of which has a mass number near 56) and one larger mass. In the case of thermal-neutron-induced fission of U235, the frequency of occurrence is observed to be greater than 1.2 x 10-6 ternary fission events per binary fission event. Ternary fission of U236: results in the formation of one light fragment (near mass 36) and two large fragments or, as in the case of Cf252, two medium fragments and one large one. These results indicate that axially asymmetric distortion modes are possible in the pre-scission configurations of the fissioning nucleus. A description is given of experiments designed to radiochemically detect the light fragment resulting from ternary fission. (author)

  20. Determination of nuclear charge resolution of a thin film detector using a 252Cf fission fragment source

    International Nuclear Information System (INIS)

    Little quantitative information is available regarding the ability of a thin film detector (TFD) to determine the nuclear charge of a low energy heavy mass ion. To compare the nuclear charge resolving power Z/ΔZ of the TFD to other heavy ion detectors, an experiment was performed where the TFD luminescence response to 252Cf fission fragments was recorded in coincidence with the gamma de-excitation of the fragments. With this technique, the TFD nuclear charge resolving power Z/ΔZ was determined to be 25.2 for the light mass fragment group. (orig.)

  1. Determination of nuclear charge resolution of a thin film detector using a [sup 252]Cf fission fragment source

    Energy Technology Data Exchange (ETDEWEB)

    Milosevich, Z.; Muga, M.L. (Florida Univ., Gainesville, FL (United States). Dept. of Chemistry); Coldwell, R.L. (Florida Univ., Gainesville, FL (United States). Dept. of Physics)

    1992-09-15

    Little quantitative information is available regarding the ability of a thin film detector (TFD) to determine the nuclear charge of a low energy heavy mass ion. To compare the nuclear charge resolving power Z/[Delta]Z of the TFD to other heavy ion detectors, an experiment was performed where the TFD luminescence response to [sup 252]Cf fission fragments was recorded in coincidence with the gamma de-excitation of the fragments. With this technique, the TFD nuclear charge resolving power Z/[Delta]Z was determined to be 25.2 for the light mass fragment group. (orig.).

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

    CERN Document Server

    Endo, S; Yoshitake, Y

    2002-01-01

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

  3. Investigation of the fission yields of the fast neutron-induced fission of {sup 233}U; Mesure de la distribution en masse et en charge des produits de la fission rapide de l'{sup 233}U

    Energy Technology Data Exchange (ETDEWEB)

    Galy, J

    1999-09-01

    As a stars, a survey of the different methods of investigations of the fission product yields and the experimental data status have been studied, showing advantages and shortcomings for the different approaches. An overview of the existing models for the fission product distributions has been as well intended. The main part of this thesis was the measurement of the independent yields of the fast neutron-induced fission of{sup 233}U, never investigated before this work. The experiment has been carried out using the mass separator OSIRIS (Isotope Separator On-Line). Its integrated ion-source and its specific properties required an analysis of the delay-parameter and ionisation efficiency for each chemical species. On the other hand, this technique allows measurement of independent yields and cumulative yields for elements from Cu to Ba, covering most of the fission yield distribution. Thus, we measured about 180 independent yields from Zn (Z=30) to Sr (Z=38) in the mass range A=74-99 and from Pd (Z=46) to Ba (Z=56) in the mass range A=113-147, including many isomeric states. An additional experiment using direct {gamma}-spectroscopy of aggregates of fission products was used to determine more than 50 cumulative yields of element with half-life from 15 min to a several days. All experimental data have been compared to estimates from a semi-empirical model, to calculated values and to evaluated values from the European library JEF 2.2. Furthermore, a study of both thermal and fast neutron-induced fission of {sup 233}U measured at Studsvik, the comparison of the OSIRIS and LOHENGRIN facilities and the trends in new data for the Reactors Physics have been discussed. (author)

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

    International Nuclear Information System (INIS)

    Safety and availability are key issues of the generation IV reactors. Hence, the three radionuclide confinement barriers, including fuel cladding, must stay tight during the reactor operation. During the primary gaseous failure, fission products xenon and krypton are released. Their fast and sensitive detection guarantees the first confinement barrier tightness. In the frame of the French ASTRID project, an optical spectroscopy technique - Cavity Ring Down Spectroscopy (CRDS) - is investigated for the gaseous fission products measurement. A dedicated CRDS set-up is needed to detect the rare gases with a commercial laser. Indeed, the CRDS is coupled to a glow discharge plasma, which generates a population of metastable atoms. The xenon plasma conditions are optimized to 110 Pa and 1.3 W (3 mA). The production efficiency of metastable Xe is then 0.8 %, stable within 0.5% during hours. The metastable number density is proportional to the xenon over argon molar fraction. The spectroscopic parameters of the strong 823.16 nm xenon transition are calculated and/or measured in order to optimize the fit of the experimental spectra and make a quantitative measurement of the metastable xenon. The CRDS is coupled to the discharge cell. The laser intensity inside the cavity is limited by the optical saturation process, resulting from the strong optical pumping of the metastable state. The resulting weak CRDS signal requires a fast and very sensitive photodetector. A 600 ppt xenon molar fraction was measured by CRDS. With the present set-up, the detection limits are estimated from the baseline noise to approximately 20 ppt for each even isotope, 60 ppt for the 131Xe and 55 ppt for the 129Xe. This sensitivity matches the specifications required for gaseous leak measurement; approximately 100 ppt for 133Xe (4 GBq/m3) and 10 ppb for stable isotopes. The odd isotopes are selectively measured, whereas the even isotopes overlap, a spectroscopic feature that applies for stable or

  5. Fission fragment charge and mass distributions in 239Pu(n ,f ) in the adiabatic nuclear energy density functional theory

    Science.gov (United States)

    Regnier, D.; Dubray, N.; Schunck, N.; Verrière, M.

    2016-05-01

    Background: Accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r process to fuel cycle optimization for nuclear energy. The need for a predictive theory applicable where no data are available, together with the variety of potential applications, is an incentive to develop a fully microscopic approach to fission dynamics. Purpose: In this work, we calculate the pre-neutron emission charge and mass distributions of the fission fragments formed in the neutron-induced fission of 239Pu using a microscopic method based on nuclear density functional theory (DFT). Methods: Our theoretical framework is the nuclear energy density functional (EDF) method, where large-amplitude collective motion is treated adiabatically by using the time-dependent generator coordinate method (TDGCM) under the Gaussian overlap approximation (GOA). In practice, the TDGCM is implemented in two steps. First, a series of constrained EDF calculations map the configuration and potential-energy landscape of the fissioning system for a small set of collective variables (in this work, the axial quadrupole and octupole moments of the nucleus). Then, nuclear dynamics is modeled by propagating a collective wave packet on the potential-energy surface. Fission fragment distributions are extracted from the flux of the collective wave packet through the scission line. Results: We find that the main characteristics of the fission charge and mass distributions can be well reproduced by existing energy functionals even in two-dimensional collective spaces. Theory and experiment agree typically within two mass units for the position of the asymmetric peak. As expected, calculations are sensitive to the structure of the initial state and the prescription for the collective inertia. We emphasize that results are also sensitive to the continuity of the collective landscape near scission. Conclusions: Our analysis confirms

  6. Fluctuations in Electronic Energy Affecting Singlet Fission Dynamics and Mixing with Charge-Transfer State: Quantum Dynamics Study.

    Science.gov (United States)

    Fujihashi, Yuta; Ishizaki, Akihito

    2016-02-01

    Singlet fission is a spin-allowed process by which a singlet excited state is converted to two triplet states. To understand mechanisms of the ultrafast fission via a charge transfer (CT) state, one has investigated the dynamics through quantum-dynamical calculations with the uncorrelated fluctuation model; however, the electronic states are expected to experience the same fluctuations induced by the surrounding molecules because the electronic structure of the triplet pair state is similar to that of the singlet state except for the spin configuration. Therefore, the fluctuations in the electronic energies could be correlated, and the 1D reaction coordinate model may adequately describe the fission dynamics. In this work we develop a model for describing the fission dynamics to explain the experimentally observed behaviors. We also explore impacts of fluctuations in the energy of the CT state on the fission dynamics and the mixing with the CT state. The overall behavior of the dynamics is insensitive to values of the reorganization energy associated with the transition from the singlet state to the CT state, although the coherent oscillation is affected by the fluctuations. This result indicates that the mixing with the CT state is rather robust under the fluctuations in the energy of the CT state as well as the high-lying CT state. PMID:26732701

  7. Mass and nuclear charge yields for sup 237 Np(2n sub th ,f) at different fission fragment kinetic energies

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, G.; Barreau, G.; Sicre, A.; Doan, T.P.; Audouard, P.; Leroux, B. (CEA Centre d' Etudes Nucleaires de Bordeaux-Gradignan, 33 - Gradignan (France)); Arafa, W.; Brissot, R.; Bocquet, J.P. (Grenoble-1 Univ., 38 (France). Inst. des Sciences Nucleaires); Faust, H. (Institut Max von Laue - Paul Langevin, 38 - Grenoble (France)); Koczon, P.; Mutterer, M. (Technische Hochschule Darmstadt (Germany, F.R.). Inst. fuer Kernphysik); Goennenwein, F. (Tuebingen Univ. (Germany, F.R.). Physikalisches Inst.); Asghar, M. (Universite des Sciences et de la Technologie Houari Boumediene, Algiers (Algeria). Inst. de Physique); Quade, U.; Rudolph, K. (Muenchen Univ. (Germany, F.R.)); Engelhardt, D. (Karlsruhe Univ. (T.H.) (Germany, F.R.)); Piasecki, E. (Warsaw Univ. (Poland))

    1990-09-03

    The recoil mass separator LOHENGRIN of the Laue-Langevin Institute Grenoble has been used to measure for the first time, the yields of light fission fragments from the fissioning system: {sub 93}{sup 239}Np; this odd-Z nucleus is formed after double thermal neutron capture in a {sub 93}{sup 237}Np target. The mass distributions were measured for different kinetic energies between 92 and 115.5 MeV, but the nuclear charge distributions were determined only up to 112 MeV. These distributions are compared to the distributions obtained for the even-even system {sub 94}{sup 240}Pu. At high kinetic energy, the mass distribution shows a prominent peak around mass number A{sub L}=106. These cold fragmentations are discussed in terms of a calculation based on a scission point model extrapolated to the cold fission case. As expected for an odd-Z fissioning nucleus, the nuclear charge distributions do not reveal any odd-even effect. The global neutron odd-even effect is found to be (8.1{plus minus}1.5)%. A simple model has been used to show that most of the neutron odd-even effect results from prompt neutron evaporation from the fragments. (orig.).

  8. Study of double phases corium atmosphere fission product time. Evolution after a nuclear reactor emergency shutdown by using Phado code

    Energy Technology Data Exchange (ETDEWEB)

    Tsilanizara, A.; Diop, C.M.; Nimal, J.C.; Nimal, B. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

    1994-12-31

    This paper deals with the PHADO code which is a part of the ESCADRE (French system of accident analysis codes for Water Reactors). The objectives of ESCADRE system is to characterize (quantitatively, qualitatively) for all the accident duration, the fission products behaviour and to define and evaluate the means for severe accident mitigation and management (limitation of core degradation and containment failure). The PHADO code treats the fission products aspects in the corium and in the atmosphere: mass, concentration, activity, residual gamma and beta decay heating for any cooling time after the emergency shutdown. (TEC).

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

    International Nuclear Information System (INIS)

    In this study, the breeding performance of a fission hybrid reactor was analyzed to provide fissile fuel for Light Water Reactors (LWR) as an alternative to the current methods of gas diffusion and gas centrifuge. LWR fuel rods containing UO2 or ThO2 fertile material were located in the fuel zone of the blanket and helium gas or Flibe (Li2BeF4) fluid was used as coolant. As a result of the analysis, according to fusion driver (D,T and D,D) and the type of coolant the enrichment of 3%-4% were achieved for operation periods of 12 and 36 months in case of fuel rods containing UO2, respectively and for operation periods of 18 and 48 months in case of fuel rods containing ThO2, respectively. Depending on the type of fusion driver, coolant and fertile fuel, varying enrichments of between 3% and 8.9% were achieved during operation period of four years

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

    International Nuclear Information System (INIS)

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

  11. Evolutionary conservation of the WASH complex, an actin polymerization machine involved in endosomal fission

    OpenAIRE

    Derivery, Emmanuel; Gautreau, Alexis

    2010-01-01

    WASH is the Arp2/3 activating protein that is localized at the surface of endosomes, where it induces the formation of branched actin networks. This activity of WASH favors, in collaboration with dynamin, the fission of transport intermediates from endosomes, and hence regulates endosomal trafficking of several cargos. We have purified a novel stable multiprotein complex containing WASH, the WASH complex, and we examine here the evolutionary conservation of its seven subunits across diverse e...

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

    Energy Technology Data Exchange (ETDEWEB)

    Goethem, G. van [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium)]. E-mail: georges.van-goethem@ec.europa.eu; Hugon, M. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium); Bhatnagar, V. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium); Manolatos, P. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium); Deffrennes, M. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium)

    2007-07-15

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

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

    International Nuclear Information System (INIS)

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

  14. A Feasibility Study on a Clean Power Fusion Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Full text: In this paper, a design concept of fusion-fission hybrid reactor for the purpose of high level radioactive waste transmutation was investigated. A concept of fusion based trans-uranium isotope (TRU) burner reactor (FTBR) was based on a low power tokamak (150 MW max) and annular ring shaped TRU core with metallic fuel (TRU 60 w/o, Zr 40 w/o) and adjacent fission product (FP) zone. Composition data for TRU and FP are assumed to be the same with those in spent fuel from 1,000 MWe PWR with 10 years decay cooling. Calculation for blanket part were performed using MCNP-X 2.6. Irradiation (burn) cycle was chosen to be 1,100 days (3 years). The power level of TRU core was set to be 2,000 MW and keff at BOC was calculated as 0.97979 and at EOC 0.85049. Calculated TBR value was 1.49 representing a self-sufficiency of fusion fuel. TRU burning was analyzed by calculating TRU mass burned per full power year (MTRU/fpy), support ratio (SR) and percentage of TRU mass burned per year (%TRU/fpy). Same parameters were also used to analyze the FP transmutation. To account for the FP produced in TRU core the net MFP/fpy and net %FP/fpy was also calculated. For toxicity analysis of long lived TRU and FP the percentage reduction of long lived inhalation toxicity (LLIhT) and long lived ingestion toxicity (LLIgT) were also calculated. MTRU/fpy was 747.11 kg with 14.25 MT of initial TRU mass loading, %TRU/fpy was 5.24% and SR was 2.24. FP mass produced in TRU core per fpy was 162.25 kg. LLIhT and LLIgT of TRU's were reduced by 9% and 6% respectively over the burn cycle. FP depletion calculations were performed for two different thicknesses of FP zone 30 cm and 50 cm to evaluate the FP loading effect on FP transmutation performance. TRU transmutation performance of FTBR was also compared with Subcritical Advance Burner Reactor (SABR) design. The comparison showed good TRU transmutation performance of FTBR with a small scaled fusion facility but it still can be improved by

  15. Neutron Beam Analysis on Materials for Nuclear Applications, Being Irradiated in Fission Reactors and Having Radioactivity

    International Nuclear Information System (INIS)

    Extensive supports are given from the public sectors to the neutron beam analysis on advanced materials developed mainly in the framework of fundamental solid state physics, through the Japan Atomic Energy Agency and the Institute for Solid State physics in University of Tokyo. However, the related activities are mainly on non-radioactive materials with some limited exceptions, though the facilities for the neutron beam analysis are installed in the radiation controlled areas. Research activities in the field of nuclear related materials have concentrated their efforts for nano structural analysis into the other techniques of the post irradiation examinations, such as the high resolution transmission microscopy, the three dimensional atom probe tomography, and the positron annihilation techniques, than the neutron beam analysis. In the meantime, more detailed analysis on the radiation induced nanostructures are becoming more and more essential for the further understanding of the radiation effects in the materials which will be used in the advanced nuclear systems, such as the nuclear fusion reactors and the generation-IV nuclear fission reactors. Utilizing of the cutting edge techniques for the nanostructural analysis on materials irradiated by neutrons, all of which cannot be installed in the limited area of available hot laboratories, is urgently demanded, of course, satisfying the related legal restrictions and the safety demands. The present study was focused on as the realization of the neutron beam analysis on the nanostructural evolutions of the superconductive materials, which will be used in the ITER, the international thermonuclear experimental reactor, being under construction in Cadarache, France, and the glassy metals, which have some unique and advantageous features for the nuclear applications. (author)

  16. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

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

    International Nuclear Information System (INIS)

    The HWR is a better candidate for large-scale transmutation of long-lived fission products. When target pins containing either Tc-99 or I-129 are positioned in the centre of each fuel bundle of a 935 MWe CANDU reactor, the transmutation half lives are 44 and 20 years, respectively, and the gross transmutation rates 60 and 48 kg/a. The positive coolant void coefficient is reduced in both cases with about 30%. When Tc-99 target pins are positioned in the moderator between the fuel bundles, the transmutation half life becomes 25 years and the gross transmutation rate 106 kg/a. This means that one HWR can serve four PWRs with equal power. The fast reactor seems most promising. When Tc-99 target pins are irradiated in moderated subassemblies in the inner core of Superphenix (∼1240 MWe), a transmutation half life of 15 years is obtained with a gross transmutation rate of 122 kg/a. These values become 18 years and 101 kg/a when non-moderated subassemblies are used for the irradiation. This implies that one fast reactor can serve four to five PWRs with equal power. The PWR seems not very effective for transmutation of Tc-99. Large inventories are needed to obtain a Tc-99 transmutation rate equal to the production rate (18 kg/a for a 900 MWe PWR). When all guide tubes of an UO2 fuelled PWR are filled with Tc-99 with density of 5 g cm-3, the transmutation half life is 39 years and the gross transmutation rate 64 kg/a. (orig./GL)

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

    International Nuclear Information System (INIS)

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

  19. Man-machine interaction at the OECD Halden reactor project

    International Nuclear Information System (INIS)

    The aims and the status of the project carried out in the Halden Man-Machine Laboratory (HAMMLAB) is presented. The capabilities and limitations of the human operator in a control room environment are investigated. (K.A.)

  20. Seismic analysis of fuelling machine support structure for CANDU6 reactor

    International Nuclear Information System (INIS)

    The fueling machine in the CANDU nuclear power plants is used to perform on-line refueling of the reactor. Canadian safety philosophy requires that the fueling machine survive the design basis earthquake. In the CANDU6 nuclear power plant there are two fueling machines, one on each side of the reactor and located in the reactor building. During reactor operation both machines can either be attached to the reactor (fueling mode) or unattached (stand-by mode). Both cases are considered for seismic qualification. The fueling machine can travel horizontally and vertically and assume any of the 380 fuel channel positions. A number of dynamic models for the fueling machine support structure are prepared using beam elements and lumped masses. Special attention is given to realistically model the linkage points between various components of the system. Spring mechanisms are represented by nonlinear spring elements in the model. The spring characteristics are determined using pull back testing of parts of the machine. These models are analyzed using multiple-level acceleration time-histories at the support points. The analysis is done using the time-history, direct integration method. PC micro computers are used to perform most of the computation work. Different routines of the STARDYNE computer program are used for that purpose. The seismic responses obtained from the analysis are used for stress analysis and verification of load ratings of components. The nonlinear time-history analysis is found to be a practical way of analyzing such a machine. The methodology, modeling techniques and results of this analysis are described in this paper

  1. The Fission Converter-Based Epithermal Neutron Irradiation Facility at the Massachusetts Institute of Technology Reactor

    International Nuclear Information System (INIS)

    A new type of epithermal neutron irradiation facility for use in neutron capture therapy has been designed, constructed, and put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). A fission converter, using plate-type fuel and driven by the MITR, is used as the source of neutrons. After partial moderation and filtration of the fission neutrons, a high-intensity forward directed beam is available with epithermal neutron flux [approximately equal to]1010 n/cm2.s, 1 eV ≤ E ≤ 10 keV, at the entrance to the medical irradiation room, and epithermal neutron flux = 3 to 5 x 109 n/cm2.s at the end of the patient collimator. This is currently the highest-intensity epithermal neutron beam. Furthermore, the system is designed and licensed to operate at three times higher power and flux should this be desired. Beam contamination from unwanted fast neutrons and gamma rays in the aluminum, polytetrafluoroethylene, cadmium and lead-filtered beam is negligible with a specific fast neutron and gamma dose, Dγ,fn/φepi [less than or approximately equal] 2 x 10-13 Gy cm2/nepi. With a currently approved neutron capture compound, boronophenylalanine, the therapeutically advantageous depth of penetration is >9 cm for a unilateral beam placement. Single fraction irradiations to tolerance can be completed in 5 to 10 min. An irradiation control system based on beam monitors and redundant, high-reliability programmable logic controllers is used to control the three beam shutters and to ensure that the prescribed neutron fluence is accurately delivered to the patient. A patient collimator with variable beam sizes facilitates patient irradiations in any desired orientation. A shielded medical room with a large window provides direct viewing of the patient, as well as remote viewing by television. Rapid access through a shielded and automatically operated door is provided. The D2O cooling system for the fuel has been conservatively designed with excess

  2. 239Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental Reactor Benchmarks

    Science.gov (United States)

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

    2014-04-01

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

  3. Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

    Science.gov (United States)

    Mushtaq, A.; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab

    2009-04-01

    Low enriched uranium foil (19.99% 235U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/ 99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.

  4. Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Low enriched uranium foil (19.99% 235U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.

  5. Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad, Federal Area 44000 (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad, Federal Area 44000 (Pakistan)

    2009-04-15

    Low enriched uranium foil (19.99% {sup 235}U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.

  6. Study of the mass, nuclear charge and kinetic energy distribution of the fission fragments produced in the reaction 237 Np (2n th, f)

    International Nuclear Information System (INIS)

    In this work, we report fission fragment mass, energy and charge distributions measured for the fissioning nucleus: 239 Np 146, This odd Z nucleus is formed after double thermal neutron capture on to the 237 Np 144 target nucleus. These measurements were performed at the I.L.L. recoil mass spectrometer ''Lohengrin'' in Grenoble. The fission fragments were registered by an ionisation chamber placed at the focal plane of the spectrometer. The obtained distributions are compared to the 240 Pu 146 fragment mass, energy and charge distributions. They are discussed within the Wilkins' scission-point model. Cold fission has been studied while selecting fragmentations with final kinetic energies close to the maximum energy released in the reaction. These cold fission events are discussed according to a calculation based on the Wilkins' scission-point model extrapolated to the cold fragmentation case. 51 refs

  7. Production of worm-gear speed reducers for NPP charging machines

    International Nuclear Information System (INIS)

    Main features and performance of three types of worm-gear speed reducers applied in WWER-1000 charging machines are presented and described. All types of reducers have a single worm; worm gearings in them are clearance free. The reducer design provides for minimal cost of production preparation and requires minimal application of special production equipment

  8. Design and analysis on tritium system of multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

    Energy Technology Data Exchange (ETDEWEB)

    Ni Muyi, E-mail: nimuyi@mail.ustc.edu.cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230026 (China); Song Yong [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230026 (China); Jin Ming; Jiang Jieqiong [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Huang Qunying [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230026 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer A concept of the tritium system was designed for the FDS-MFX. Black-Right-Pointing-Pointer The system parameters were presented and discussed in detail. Black-Right-Pointing-Pointer A theoretical analysis of tritium recovery system has been made on the operation condition. Black-Right-Pointing-Pointer Three step TEP system was design and its permeating capacity was estimated. Black-Right-Pointing-Pointer The model of three-column ISS and the SDS was also carried out. - Abstract: As early application of fusion technology, the fusion-fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion-fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.

  9. Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    Highlights: ► A concept of the tritium system was designed for the FDS-MFX. ► The system parameters were presented and discussed in detail. ► A theoretical analysis of tritium recovery system has been made on the operation condition. ► Three step TEP system was design and its permeating capacity was estimated. ► The model of three-column ISS and the SDS was also carried out. - Abstract: As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.

  10. Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961)

    International Nuclear Information System (INIS)

    - Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott 'β' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Release of fission product species from UO2, and to a limited extent from (U, Pu)02 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO2 powder 20% 235U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x1016 neutrons m-2s-1 providing a heat rating within the samples of 34.5 MW/teU

  13. Modeling of constituent redistribution and fission product migration in fast reactor U-Pu-Zr fuel

    International Nuclear Information System (INIS)

    Radial constituent redistribution in a fast reactor U-Pu-Zr fuel is an important phenomenon that occurs because the fuel alloy has thermal gradients and poly-phase fields at the typical operation temperature. In a typical temperature range (500-700degC), Zr moves from the mid-radius region to the fuel center region and the fuel surface region. Because of this phenomenon, the homogeneous fuel at beginning of life turns into locally heterogeneous fuel. Most of the thermophysical properties change locally, as does fuel performance. Fuel constituent redistribution of U-Pu-Zr is modeled by treating Pu as sessile element and therefore by assuming a pseudo-binary system. Fission product lanthanides (LA) migrate to the fuel surface. LA migration appears to be due both to direct vapor transport and diffusion through the fuel matrix. Large pores at the low Zr zone and fuel periphery may support for LA precipitates. LA diffusion through Pu also contributes to observed LA migration. Because Pu is relatively sessile, however, LA migration by diffusion through the fuel matrix is relatively small. Upon migration to the fuel surface, LA and Pu react with Fe-base alloy cladding such as HT9 and D9 whereas U and Zr do not. The LA and Pu reaction with cladding is via interdiffusion. (author)

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

    International Nuclear Information System (INIS)

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

  15. Conceptual Analysis of the Economic Feasibility of Fission Electric Cell Reactors

    International Nuclear Information System (INIS)

    The United States Department of Energy, Nuclear Energy Research Initiative (NERI) Direct Energy Conversion (DEC) project began in August of 1998 with the goal of developing a direct energy conversion process suitable for commercial development. With roughly two thirds of the project completed, we believe a viable direct energy device could be economic. This paper describes the financial basis behind that belief for one proposed DEC reactor, the magnetically insulated fission electric cell (FEC). It also illustrates the value of economic analysis even in these early phases of a research project. The financial basis consists of a conceptual level Economic Model comprised of five modules. The Design Model provides technical specification to other modules. The Fuel Cost Model estimates fuel expenses based on current spot market prices applied over a wide range of fuel enrichment. The Operating Cost Model uses published correlations to provide rough order of magnitude non-fuel operating costs. The Capital Cost model uses analogy and parametric estimating techniques to generate capital cost estimates for a DEC power plant. Finally, the financial model combines output from the other models to produce a Net Present Value analysis with cost of generation as the independent variable. Model results indicate that several FEC geometric configurations could be economic. Within these configurations, optimums exist. Finally, the model demonstrates that the most efficient design is not necessarily the most economic. (authors)

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

    International Nuclear Information System (INIS)

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

  17. Vapor transport of fission products in postulated severe light water reactor accidents

    International Nuclear Information System (INIS)

    A methodology based on chemical thermodynamics has been developed to treat the transport of volatile fission products (FPs) through the core and the primary system. The FPs considered are cesium, iodine, tellurium, strontium, and ruthenium, which may pose the major biohazard in postulated severe accidents in light water reactors. The vapor transport of FPs depends on the volatilities of the chemical compounds that are formed in the carrier gas environment in which the FPs are released and transported. Chemically stable forms were evaluated by minimizing the total free energies of the FP/ fuel/gas environment systems. Many gaseous species for each FP were considered and their partial pressures calculated over a range of temperatures (600 to 3000K), the carrier gas environments (total pressure and ratio of H2/H2O), and the total amount of FPs in the system. It was found that the major dependence of the concentration of the FPs was on the gas temperature, and a model was developed to predict the source of volatile FPs. The model showed that the FPs leaving the core region would condense in the cooler regions of the upper plenum and/or the primary system either on the cold surfaces or be transported further as aerosols

  18. Nuclear Data Requirements for the Production of Medical Isotopes in Fission Reactors and Particle Accelerators

    CERN Document Server

    Garland, M A; Talbert, R J; Mashnik, S G; Wilson, W B

    1999-01-01

    Through decades of effort in nuclear data development and simulations of reactor neutronics and accelerator transmutation, a collection of reaction data is continuing to evolve with the potential of direct applications to the production of medical isotopes. At Los Alamos the CINDER'90 code and library have been developed for nuclide inventory calculations using neutron-reaction (En < 20 MeV) and/or decay data for 3400 nuclides; coupled with the LAHET Code System (LCS), irradiations in neutron and proton environments below a few GeV are tractable; additional work with the European Activation File, the HMS-ALICE code and the reaction models of MCNPX (CEM95, BERTINI, or ISABEL with or without preequilibrium, evaporation and fission) have been used to produce evaluated reaction data for neutrons and protons to 1.7 GeV. At the Pacific Northwest National Laboratory, efforts have focused on production of medical isotopes and the identification of available neutron reaction data from results of integral measuremen...

  19. Fission product iodine release and retention in nuclear reactor accidents— experimental programme at PSI

    Science.gov (United States)

    Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

    2003-01-01

    Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

  20. Overview of research by the fission group in Trombay

    Indian Academy of Sciences (India)

    R K Chourdhury

    2015-08-01

    Nuclear fission studies in Trombay began nearly six decades ago, with the commissioning of the APSARA research reactor. Early experimental work was based on mass, kinetic energy distributions, neutron and X-ray emission in thermal neutron fission of 235U, which were carried out with indigenously developed detectors and electronics instrumentation. With the commissioning of CIRUS reactor and the availability of higher neutron flux, advanced experiments were carried out on ternary fission, pre-scission neutron emission, fragment charge distributions, quarternary fission, etc. In the late eighties, heavy-ion beams from the pelletron-based medium energy heavy-ion accelerator were available, which provided a rich variety of possibilities in nuclear fission studies. Pioneering work on fragment angular distributions, fission time-scales, transfer-induced fission, -ray multiplicities and mass–energy correlations were carried out, providing important information on the dynamics of the fission process. More recently, work on fission fragment -ray spectroscopy has been initiated, to understand the nuclear structure aspects of the neutron-rich fission fragment nuclei. There have also been parallel efforts to carry out theoretical studies in the areas of shell effects, superheavy nuclei, fusion–fission dynamics, fragment angular distributions, etc. to complement the experimental studies. This paper will provide a glimpse of the work carried out by the fission group at Trombay in the above-mentioned topics.

  1. SOFIA, a Next-Generation Facility for Fission Yields Measurements and Fission Study. First Results and Perspectives

    Science.gov (United States)

    Audouin, L.; Pellereau, E.; Taieb, J.; Boutoux, G.; Béliera, G.; Chatillon, A.; Ebran, A.; Gorbinet, T.; Laurent, B.; Martin, J.-F.; Tassan-Got, L.; Jurado, B.; Alvarez-Pol, H.; Ayyad, Y.; Benlliure, J.; Caamano, M.; Cortina-Gil, D.; Fernandez-Dominguez, B.; Paradela, C.; Rodriguez-Sanchez, J.-L.; Vargas, J.; Casarejos, E.; Heinz, A.; Kelic-Heil, A.; Kurz, N.; Nociforo, C.; Pietri, S.; Prochazka, A.; Rossi, D.; Schmidt, K.-H.; Simon, H.; Voss, B.; Weick, H.; Winfield, J. S.

    2015-10-01

    Fission fragments play an important role in nuclear reactors evolution and safety. However, fragments yields are poorly known : data are essentially limited to mass yields from thermal neutron-induced fissions on a very few nuclei. SOFIA (Study On FIssion with Aladin) is an innovative experimental program on nuclear fission carried out at the GSI facility, which aims at providing isotopic yields on a broad range of fissioning systems. Relativistic secondary beams of actinides and pre-actinides are selected by the Fragment Separator (FRS) and their fission is triggered by electromagnetic interaction. The resulting excitation energy is comparable to the result of an interaction with a low-energy neutron, thus leading to useful data for reactor simulations. For the first time ever, both fission fragments are completely identified in charge and mass in a new recoil spectrometer, allowing for precise yields measurements. The yield of prompt neutrons can then be deduced, and the fission mechanism can be ascribed, providing new constraints for fission models. During the first experiment, all the technical challenges were matched : we have thus set new experimental standards in the measurements of relativistic heavy ions (time of flight, position, energy loss).This communication presents a first series of results obtained on the fission of 238U; many other fissioning systems have also been measured and are being analyzed presently. A second SOFIA experiment is planned in September 2014, and will be focused on the measurement of the fission of 236U, the analog of 235U+n.

  2. FFTF (FAST FLUX TEST FACILITY) REACTOR CHARACTERIZATION PROGRAM ABSOLUTE FISSION RATE MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    FULLER JL; GILLIAM DM; GRUNDL JA; RAWLINS JA; DAUGHTRY JW

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  3. FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A.; Daughtry, J.W.

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  4. N-reactor charge-discharge system analysis

    International Nuclear Information System (INIS)

    This report documents an analysis of the existing systems in the N-Reactor fuel flow path. It recommends equipment improvements and changes in that path to allow the charge-discharge rates to be increased to 500 tubes per outage without increasing reactor outage time. The estimated program cost of $14 million is projected over an estimated 3-year period. It does not include costs detailed as part of the existing restoration program or any costs that are considered as normal maintenance. The recommendations contained in this report provide a direction and goal for every critical aspect of the fuel flow path. The way in which these recommendations are implemented may greatly affect the schedule and costs. Previous studies by UNC have shown that enhanced fuel element handling has the potential of increasing productivity by 33 days at a cost benefit estimated at $18 million per year. Enhanced fuel handling provides the greatest potential for productivity improvement of any of the areas considered in these studies

  5. Human machine interaction research experience and perspectives as seen from the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    In this paper a short review is given on important safety issues in the field of human machine interaction as expressed by important nuclear organisations such as USNRC, IAEA and the OECD NEA. Further on, a presentation is offered of research activities at the OECD Halden Reactor Project in the field of human machine interaction aiming to clarify some of the issues outlined by the above mentioned organisations. The OECD Halden Reactor Project is a joint undertaking of national nuclear organisations in 19 countries sponsoring a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency. One of the research areas is the man-machine systems research addressing the operator tasks in a control room environment. The overall objective is to provide a basis for improving today's control rooms through introduction of computer-based solutions for effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. (author)

  6. Benchmark analysis of fission-rate distributions in a series of spherical depleted-uranium assemblies for hybrid-reactor design

    International Nuclear Information System (INIS)

    Highlights: • We do simulations using MCNP code and ENDF/B-V.0 library. • The fission rate distribution on depleted uranium assemblies was analyzed. • The calculations overestimate the measured fission rates. • The observed differences are discussed. - Abstract: The nuclear performance of a fission blanket in a hybrid reactor has been validated by analyzing fission-rate experiments with a series of spherical depleted-uranium assemblies. Calculations were made with the Monte–Carlo transport code MCNP5 and the ENDF/B-V.0 continuous-energy cross sections and compared to the measured results. The ratios of calculated to experimental values (C/E) for the fission rate and the fission-rate ratio of 238U to 235U are presented along with a discussion of the validation of the ENDF/B-V.0 library regarding its use in the design of the fission blanket. Overestimations are observed in the calculation of the 238U and 235U fission rates at all positions, except the ones near the outer surfaces of the assemblies, and the C/Es of the fission rate decreased as the thickness of the depleted-uranium (DU) layer increased, while most of the C/Es of the fission-rate ratio of 238U to 235U were close to unity, being within the range of 0.95–1.05

  7. The behaviour of fission products in the HTGR fuel irradiated in the IVV-2M reactor

    International Nuclear Information System (INIS)

    The results of the post-irradiation investigations of fission products behaviour in HTGR fuel and its main elements such as kernels, protective coatings and matrix graphite are considered. The dominating role of SiC layer in the protective coating of coated particles in the retention of the volatile and solid fission products, being of great radiological importance, is noticed. (author)

  8. Development of a high temperature, high sensitivity fission counter for liquid metal reactor in-vessel flux monitoring

    International Nuclear Information System (INIS)

    Advanced liquid metal reactor concepts such as the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Inherently Safe Module (PRISM) have relatively large pressure vessels that necessitate in-vessel placement of the neutron detectors to achieve adequate count rates during source range operations. It is estimated that detector sensitivities of 5 to 10 counts/center dot/s/center dot//sup /minus/1//center dot/[neutron/(cm2/center dot/s)]/sup /minus/1/ will be required for the initial core loading. The Instrumentation and Controls Division of Oak Ridge National Laboratory has designed and fabricated a fission counter to meet this requirement which is also capable of operating in uncooled instrument thimbles at primary coolant temperatures of 500 to 600/degree/C. Components are fabricated from Inconel-600, and high temperature alumina insulators are employed. The transmission line electrode configuration is utilized to minimize capacitive loading effects

  9. In-pile fission gas release and swelling of UO2 fuel in light water reactors

    International Nuclear Information System (INIS)

    A theoretical model is proposed to describe the release of in-pile fission gases from UO2 fuel. It is assumed that the gas release takes place by migrating of the fission gas in the grains and on the grain boundaries to the grain edge tunnels. The release of fission gas is modelled taking into account the following physical processes: fission gas generation in the fuel matrix, bubble nucleation and coalescence, atomic and bubble migration, irradiation induced resolution from intra- and intergranular bubbles, and grain boundary bubbles growth and interlinkage. In addition to the diffusion process, the grain boundary sweeping mechanism involving equiaxed grain growth is also considered. The mechanism of fission gas release at low temperature is not considered here. The diffusion equations set is solved numerically and the theoretical results are compared with the experimental data. (author)

  10. An Evaluation on the Influence of Axial Reflector Thickness into the Fission Source Convergence in MC Eigenvalue Calculation of VHTR Prismatic Reactor

    International Nuclear Information System (INIS)

    Analyses of the prismatic VHTR with Monte Carlo method suffer from slow fission source convergence. MHTGR-350 is a prismatic VHTR, which has an asymmetric reflector thickness along the axial direction. In this case, fission source distribution also becomes strong asymmetrical distribution according to the asymmetric reactor reflector thickness. Therefore, the converged fission source must be verified to pursue the Monte Carlo simulation of the reactor type. In this study, how the axial reflector thickness affects the fission source convergence was evaluated with changing the prismatic VHTR reflector thickness. In this study, how the axial reflector thickness affects the fission source convergence was evaluated. For the symmetric reflector cases, the results show the fission source distribution was converged within 60th cycle. However, in the cases of the asymmetric reflector thickness, it is notified that the convergence cycle of the fission source distribution exceeded 200th cycle. Analysis shows that the inactive cycle for the Monte Carlo eigenvalue calculation should be considerably decided when the reactor has asymmetric reflector thicknesses such as the MHTGR-350. It is expected that these results can be directly used for evaluating and analyzing the prismatic VHTR with Monte Carlo method

  11. History and Actual State of Non-HEU Fission-Based Mo-99 Production with Low-Performance Research Reactors

    Directory of Open Access Journals (Sweden)

    S. Dittrich

    2013-01-01

    Full Text Available Fifty years ago, one of the worldwide first industrial production processes to produce fission-Mo-99 for medical use had been started at ZfK Rossendorf (now: HZDR, Germany. On the occasion of this anniversary, it is worth to mention that this original process (called LITEMOL now together with its target concept used at that time can still be applied. LITEMOL can be adapted very easily to various research reactors and applied at each site, which maybe still of interest for very small-scale producers. Besides this original process, two further and actually proven processes are suitable as well and recommended for small-scale LEU fission Mo-99 production also. They are known under the names KSA/KSS COMPACT and ROMOL LITE and will be described below.

  12. TRANCS, a computer code for calculating fission product release from high temperature gas-cooled reactor fuel, (2)

    International Nuclear Information System (INIS)

    This report describes the calculation procedure of the TRANCS code, which deals with fission product transport in fuel rod of high temperature gas-cooled reactor (HTGR). The fundamental equation modeled in the code is a cylindrical one-dimensional diffusion equation with generation and decay terms, and the non-stationary solution of the equation is obtained numerically by a finite difference method. The generation terms consist of the diffusional release from coated fuel particles, recoil release from outer-most coating layer of the fuel particle and generation due to contaminating uranium in the graphite matrix of the fuel compact. The decay term deals with neutron capture as well as beta decay. Factors affecting the computation error has been examined, and further extention of the code has been discussed in the fields of radial transport of fission products from graphite sleeve into coolant helium gas and axial transport in the fuel rod. (author)

  13. Migration behaviour of fission products in and from spherical HTR fuel elements (SAPHIR loop experiments in the PEGASE reactor)

    International Nuclear Information System (INIS)

    The diffusion behaviour of some metallic fission products was evaluated for the irradiation experiments SAPHIR 4-10 and 11 performed by KFA Juelich in the PEGASE reactor, Cadarache (France). Diffusion coefficients for cesium in PyC coatings and for cesium and silver in graphitic fuel element matrix were obtained by analysing the fission product concentration profiles. The reason for the observed different diffusion behaviour of Cs 134 and Cs 137 is discussed as well as that for elevated concentrations of cesium, silver, and ruthenium near the fuel element surface. A simple model for assessing the trapping of metals in a graphite matrix is introduced. Release from the fuel element surfaces into the surrounding gas was found to be governed by direct recoil at low temperatures. (orig.)

  14. Practical limitations for the release of fission products during the operation of a research reactor: a case study of BR2

    Energy Technology Data Exchange (ETDEWEB)

    Joppen, F. [Health Physics and Safety Department, SCK-CEN, B-2400 Mol (Belgium)

    1998-07-01

    Failure of the cladding of a fuel element is an event occurring from time to time while operating a research reactor. As a consequence, fission products are released in the primary circuit of the reactor. This contamination means no direct hazard for the workers or for the environment in case the reactor has a closed primary circuit. The operator can decide to continue the irradiation to finish a scientific experiment or a commercial isotope production program. However, the operator cannot prolong the cycle regardless the concentration fission products in the primary loop. Beside the limitations imposed by the regulatory authorities, ALARA considerations should be taken into account. An untimely stop of the reactor can have serious financial consequences and prolonged operation causes higher radiation doses. This paper gives an overview of decision process applied in case of detection of fission products in the primary circuit of BR2. (author)

  15. Practical limitations for the release of fission products during the operation of a research reactor: a case study of BR2

    International Nuclear Information System (INIS)

    Failure of the cladding of a fuel element is an event occurring from time to time while operating a research reactor. As a consequence, fission products are released in the primary circuit of the reactor. This contamination means no direct hazard for the workers or for the environment in case the reactor has a closed primary circuit. The operator can decide to continue the irradiation to finish a scientific experiment or a commercial isotope production program. However, the operator cannot prolong the cycle regardless the concentration fission products in the primary loop. Beside the limitations imposed by the regulatory authorities, ALARA considerations should be taken into account. An untimely stop of the reactor can have serious financial consequences and prolonged operation causes higher radiation doses. This paper gives an overview of decision process applied in case of detection of fission products in the primary circuit of BR2. (author)

  16. Study of the emission of a light particle charged during the fission of 235U by thermal neutron

    International Nuclear Information System (INIS)

    In a first part, this research thesis discusses the existing theories of the mechanism of emission of light particles charged of tri-partition (tri-partition is defined as an event involving two big fragments of masses comparable with those obtained in binary fission, and a charged light particle). Then, the author presents and reports an experiment performed by suing nuclear emulsions. Another type of experiment is then presented which allows the measurement of masses and energies of tri-partition fragments. The author then presents theoretical calculations which have been performed in order to find again some characteristics of tri-partition. These calculations are mainly based on Coulomb repulsion between various fragments

  17. MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(Ei), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after

  18. Modeling of a double fission chamber using MCNPX for power calibration at the zero-power teaching reactor CROCUS

    International Nuclear Information System (INIS)

    MCNPX-2.5 simulations and experiments were performed to improve the power prediction of the zero-power teaching reactor CROCUS at the Ecole Polytechnique Federale de Lausanne (EPFL) using a calibrated double fission chamber (DFC). The CROCUS facility is a zero-power critical reactor used for educational purposes. Traditionally, the core power is determined by irradiating thin gold foils placed along the core centre and by measuring the 411 keV γ-rays on HPGe detectors. The average 197Au(n,γ) self-shielded macroscopic cross-section obtained with the deterministic BOXER code (1σ - 10%) is employed to determine the flux and the reactor power. To benchmark the BOXER calculations, a DFC containing known amounts of enriched 235U and 239Pu deposits was installed within the reflector core and simulated with MCNPX-2.5/JEF-2.2. Particular care was taken to model the fissile deposits allowing to reduce the power uncertainty to 2% compared to the gold foil technique. A code-to-code comparison (BOXER vs. MCNPX) was performed and the results have shown a good agreement (2 to 5%) for most of the quantities calculated (flux, reaction rates). However, the normalization factor differed by 17% (flux-to-power ratio). Consequently, the core power was overestimated by 17% until now. Finally, the current investigations lead to an improved fission power determination and contribute to better core safety standard. (author)

  19. Preliminary Neutronics Calculation of Thorium-Based and M A Transmutation Breeding Blanket for Hybrid Fusion-Fission Reactor

    International Nuclear Information System (INIS)

    Hybrid fusion-fission reactor has advantages of production of nuclear fuel and transmutation of long-life nuclear waste and having inherent safety, at the same time, demand is significantly reduced compare to the pure fusion reactor. Breeding blanket is the key part of the fusion-fission reactor and in the past, the uranium-plutonium blanket concept was widely investigated. Considering the problem of uranium-plutonium cycle and abundant in thorium in our country, in this work,a thorium-based breeding and MA (minor actinides) transmutation blanket concept was proposed and the preliminary neutronics calculation was discussed. One-dimensional transport and burnup calculation code BISONC and Monte-Carlo transport code MCNP were used to calculate the key parameters, such as tritium breeding ratio, production of 233U mass and power density,and so on. The fuel of 233U enrichment can be 3.65%. It is the foundation for optimization of the blanket. (authors)

  20. An analytical assessment of the chemical form of fission products during postulated severe accidents in SRS production reactors

    International Nuclear Information System (INIS)

    An analysis has been performed to determine the principal chemical forms for the structural and fission product elements during a postulated severe core damage accident in a tritium-producing core in the Savannah River Site (SRS) reactors. These reactors are powered with UAlx fuel. Six core elements, cesium, iodine, tellurium, strontium, barium, and lithium, were emphasized in this analysis. Other elements also included were aluminum, hydrogen, oxygen, uranium, molybdenum, silicon, zirconium, magnesium, iron, chromium, nickel, cadmium, zinc, copper, manganese, nitrogen, and argon. The masses of each of the constituents used in the analysis were based on end-of-core-life masses for the fuel, structural, and fission product elements and on core gas volume, temperature, and pressure for steam nitrogen and argon. A chemical equilibrium analysis was performed using the Facility for Analysis of Chemical Thermodynamics (FACT) computer code at three temperatures (800, 1,100, and 1,400 K) and two pressures (1 and 10 atm). These temperatures and pressures are typical for postulated severe core accidents in the Advanced Test Reactor

  1. a comparative analysis of fission gas diffusion models in pressurized water reactor fuels

    International Nuclear Information System (INIS)

    Calculation of fission gas release has a great importance in thermal and mechanical analysis of nuclear fuel. In general, gas release calculations have been carried out with mechanical or probabilistic gas release models which have details in microstructure relations at different level.In this study, a standard PWR fuel at different power levels and burnups was modeled with FRAPCON-2 computer program. Four fission gas release models: Beyer-Hann, ANS5-4, MacDonald-Weisman, and Grass were used in fission gas release calculation. Results were evaluated on the basis of fuel parameters such as fuel temperature, fuel rod gas increase, fuel rod gas pressure and gas release fraction

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

    International Nuclear Information System (INIS)

    In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O2-kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 16000C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP)

  3. Fission chambers as the best suited detector family for in vessel neutron instrumentation of the sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    The sodium-cooled fast reactor (SFR) is one of the main options as a next generation reactor. Its in-vessel instrumentation is required to detect any abnormal situation at a sufficiently early stage, and thus to deliver measurements that are reliable and easy to interpret over several reactor cycles. In this paper we pick up the detector family that is the best suited for the in-vessel SFR instrumentation with respect to this requirement. Three types of detectors that are widely used for in-core neutron measurements are reviewed: fission chambers, boron-lined proportional counters and self-powered neutron detectors. We use as an input data neutron and gamma spectra that have been computed for a preliminary design of the SFR in different locations. We compute for each detector family the expected signal, to assess whether its level is sufficient. The evolution of the signal due to the depletion of the active part of the detectors is also addressed, to examine whether it is compatible with long term measurements. The issue of leakage current and irradiation damages is examined. Fission chambers appear to be the best suited detector family for in vessel neutron instrumentation of the SFR, able to deliver an interpretable signal for a wide dynamic of reactor power and for three or more operating cycles. This conclusion is supported by three key assets: the possibility to choose the fissile deposit according to the location in the reactor, the excellent rejection of the gamma events by using the Campbelling mode, and the intrinsic wide dynamic of those detectors when combining pulse mode and Campbelling mode. The Campbelling mode is also a convenient way to deal with the leakage current. In contrast, the two other types are shown to be inadequate for SFR measurements. (authors)

  4. Thermal-hydraulics deisgn and analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    A conceptual design of fusion-fission hybrid reactor for energy production, named FDS-EM (Energy Multiplier), was proposed. It was preliminary designed to generate about an electricity power of about 1.0 GW with self-sustaining tritium cycle. This contribution performed the thermal-hydraulics design and analyses for FDS-EM water-cooled blanket. The typical thermal-hydraulics parameters were designed by using mature technologies of PWR, and temperature and stress analyses were carried out, according to typical parameters of the blanket. The results preliminarily demonstrated the engineering feasibility of the design. (authors)

  5. KINAX-3, 1-D 1 Group Reactor Kinetics with Xe and I and Fission Products Heating and Auto-Control

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: Studies the kinetic behaviour of a power reactor, using one-dimensional, one neutron energy group model. Includes effects of Xenon, Iodine, fission product heating, auto-control, and other moving absorbers. 2 - Method of solution: Semi-implicit solution in time (Crank-Nicolson). Simple three point approximation to neutron diffusion terms in equations. 3 - Restrictions on the complexity of the problem: Maximum number of mesh points - 50; Maximum number of delay groups - 6; Maximum number of material temperatures calculated at each point - 17

  6. The thermal column. A new irradiation position for fission-track dating in the University of Pavia Triga Mark II nuclear reactor

    International Nuclear Information System (INIS)

    In the present paper a new irradiation position arranged for fission-track dating in the Triga Mark II reactor of the University of Pavia is described. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively high neutron thermalization (cadmium ratio of 85.3 for gold and 643 for cobalt) and lack of significant fluence spatial gradients are very favorable factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author). 9 refs., 2 figs., 4 tabs

  7. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDUR 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 233U isotope which has very high quality fission cross-section with thermal neutrons. 233U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2D + 3T →? 4He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li2BeF4, LiF-NaF-BeF2, Li20Sn80, natural Lithium and Li17Pb83, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li3N, Li2O, Li2O2, Li2TiO3, Li4SiO3, Li2ZrO3, LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S8-P3 approximation with Gaussian quadratures using

  10. Evaluation of the damage resulting from fission fragments in the fuel elements of the ETRR-2 materials testing reactor

    International Nuclear Information System (INIS)

    The damage resulting from fission fragments in the fuel elements of the ETRR-2 Materials Testing Reactor (MTR) was evaluated using the TRIM-86 computer code. The effect of very low yield fission ions (Zn and Gd) and high yield ions (Mo and La) on the structural changes of the aluminum matrix for the enriched U3O8 dispersed fuel was studied. The kinetic energy of the highly ionized fission fragments Zn, Mo, La and Gd was evaluated to be 106. 93. 61 and 48 MeV respectively. The tracks and distributions of ions and recoils were recorded and the mean ranges were calculated. The recoil and vacancy distributions as a function of track depth was studied mainly for the high yield Mo and La fission ions. The results indicated that the average number of recoils and vacancies produced from one ion of Mo and/or La amounted to 367x104 for recoils and 3.61 x 104 and 6.55 x 104 for vacancies, respectively. The replacement collisions amounted to 4% and 1% for Mo and La ions. The Total energy to aluminum recoils resulting from Mo and La ions was 3.67 and 6.77 MeV. The corresponding energies per recoil are 82 eV and 100 eV. The displacement energy required to create vacancy-recoil pair was found to be 20 5 eV and and the minimum energies with recoils to create Frenkle pairs are 102.5 and 120.5 eV for Mo and La ions

  11. JRC's on-line fission gas release monitoring system in the high flux reactor Petten

    International Nuclear Information System (INIS)

    For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed, dubbed the “Sweep Loop Facility” (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experiments. The SLF enables continuous and independent surveillance of all gas circuits. The release of volatile fission products of the in-pile experiments is monitored by continuous gas purging. The fractional release of these fission products, defined as the ratio between release rate of a gaseous fission isotope (measured) to its instantaneous birth rate (calculated), is a licensing-relevant test for HTR fuel. The newly developed gamma spectrometry station allows for higher measurement frequencies, thus enabling follow-up of rapid and massive release transients. The designed stand-alone system was tested and fully used through the final irradiation period of the HFR-EU1 experiment which was terminated on 18 February 2010. Its robustness allowed us to use it as extra safety instrumentation. This paper describes the gas activity measurement technique based on HPGe gamma spectrometry and illustrates how qualitative and quantitative analysis of volatile fission products can be performed on-line.

  12. Technological research on Recycling of Actinides and fission products (RAS). Irradiations in the High Flux Reactor (HFR), Petten, Netherlands

    International Nuclear Information System (INIS)

    The purpose of the title irradiations is to study the efficiency and technical feasibility of possible transmutation processes for those long-lived actinides and fission products, that contribute to long-term radiotoxicity and leaking risks of geological storage. A cooperative research program (EFFTRA or Experimental Feasibility of Targets for TRAnsmutation) has been set up for irradiations of technetium, iodine and americium in the thermal reactor HFR and the fast reactor Phenix. A radiation program for fission products is in progress in the HFR. An inert matrix concept is developed, in which the actinide is mixed with a ceramic material, which hardly reacts with neutrons and actinides and containment materials. Irradiation experiments with candidate inert matrices will be carried out in the HFR. Also, the feasibility of transmutation of americium in a thermal spectrum will be demonstrated by means of a long-range experiment in the HFR. Plans are elaborated for the irradiation of plutonium in inert matrices in the HFR to realize an efficient transmutation of existing supplies, both military and civil, of plutonium. 8 figs., 4 tabs., 18 refs

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

    International Nuclear Information System (INIS)

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

  14. Effects of T-odd asymmetry of the emission of light charged particles and photons during fission of heavy nuclei by polarized neutrons

    International Nuclear Information System (INIS)

    The new physical effects of T-odd asymmetry of the emission of light charged particles (LCPs) during the ternary fission of some heavy nuclei by cold polarized neutrons have been experimentally studied. The coefficients of triple scalar and vector correlation of the pulses of light particles and fission fragments (TRI effect) and the fivefold correlation of the same vectors (ROT effect) have been measured. These effects are believed to be caused by the rotation of polarized fissioning system around its polarization direction. The treatment of the experimental data for LCPs in the framework of this hypothesis leads to a good agreement between the calculation results and experimental data. The calculated value of the angle of rotation of the fission axis in the ternary fission of the polarized fissioning 236U* compound nucleus was used to process the results of measuring the ROT effect for γ photons from binary-fission fragments of the same nucleus. A satisfactory description of these experimental data is obtained which serves a convincing confirmation of the rotation hypothesis.

  15. A Monte Carlo simulation of a simplified reactor by decomposition of the neutron spectrum into fission, intermediate and thermal distributions

    Energy Technology Data Exchange (ETDEWEB)

    Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T. de, E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: vilhena@mat.ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre (Brazil). Grupo de Estudos Nucleares. Escola de Engenharia; Leite, Sergio Q. Bogado, E-mail: sbogado@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    In this paper the neutron spectrum of a simulated hypothetical nuclear reactor is decomposed as a sum of three probability distributions. Two of the distributions preserve shape with time but not necessarily the integral. One of the two distributions is due to fission, i.e. high neutron energies and the second a Maxwell-Boltzmann distribution for low (thermal) neutron energies. The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. This procedure is effective in attaining two objectives, the first is to include effects due to up-scattering of neutrons, and the second is to optimize computational time of the stochastic method (tracking and interaction). The simulation of the reactor is done with a Monte Carlo computer code with tracking and using continuous energy dependence. This code so far computes down-scattering, but the computation of up-scattering was ignored, since it increases significantly computational processing time. In order to circumvent this problem, one may recognize that up-scattering is dominant towards the lower energy end of the spectrum, where we assume that thermal equilibrium conditions for neutrons immersed in their environment holds. The optimization may thus be achieved by calculating only the interaction rate for neutron energy gain as well as loss and ignoring tracking, i.e. up-scattering is 'simulated' by a statistical treatment of the neutron population. For the fission and the intermediate part of the neutron spectrum tracking is taken into account explicitly, where according to the criticality condition the integral of the fission spectrum may depend on time. This simulation is performed using continuous energy dependence, and as a rst case to be studied we assume a recurrent regime. The three calculated distributions are then used in the Monte Carlo code to compute the subsequent Monte Carlo steps with subsequent updates

  16. Fission product transport in the primary system of a pebble bed high temperature reactor with direct cycle

    International Nuclear Information System (INIS)

    Transport and deposition of fission products in the primary system of a small pebble bed high temperature reactor with directly coupled gas turbine have been investigated. The reactor has a thermal power of 40 MW and is intended for heat and power cogeneration. Four radionuclides have been identified as most relevant because of volatility and radiotoxicity: 137Cs, 90Sr, 110mAg, 131I. With the code PANAMA the fraction of failed coated fuel particles has been calculated. The diffusion of the fission products to the fuel element outer surface has been calculated with the FRESCO code. Transport and deposition of the fission products within the primary system has been analysed with the code MELCOR. Under normal operating conditions the release rate of the short-lived 131I reaches a constant level rather quickly, contrary to the longer lived 137Cs and 90Sr which show a steady increase of the release rate during burn-up. Under incident conditions the retention capability of the fuel elements' graphite is strongly reduced. The release from the intact coated particles remains negligible compared to the release from the defect coated particles. After a ten year operation period, the total activity of the released nuclides in the primary system is about 58 GBq. The highest activity is found in the pre-cooler. Other components with high activities are the recuperator and the compressor. These components are contaminated mainly by 110mAg. The gas ducts in the energy conversion unit are contaminated by 110mAg and 131I. Contamination as a consequence of incident conditions is difficult to estimate, because it depends on a number of phenomena. Under the assumption that 10 fuel elements are damaged, the activity is about 44 GBq. (author)

  17. Modeling of the saturation current of a fission chamber taking into account the distorsion of electric field due to space charge effects

    CERN Document Server

    Poujade, O; Poujade, Olivier; Lebrun, Alain

    1999-01-01

    Fission chambers were first made fifty years ago for neutron detection. At the moment, the French Atomic Energy Commission \\textsf{(CEA-Cadarache)} is developing a sub-miniature fission chamber technology with a diameter of 1.5 mm working in the current mode (Bign). To be able to measure intense fluxes, it is necessary to adjust the chamber geometry and the gas pressure before testing it under real neutron flux. In the present paper, we describe a theoretical method to foresee the current-voltage characteristics (sensitivity and saturation plateau) of a fission chamber whose geometrical features are given, taking into account the neutron flux to be measured (spectrum and intensity). The proposed theoretical model describes electric field distortion resulting from charge collection effect. A computer code has been developed on this model basis. Its application to 3 kinds of fission chambers indicates excellent agreement between theoretical model and measured characteristics.

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

    International Nuclear Information System (INIS)

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

  19. Radioactive Beams from 252CF Fission Using a Gas Catcher and an ECR Charge Breeder at ATLAS

    CERN Document Server

    Pardo, Richard C; Hecht, Adam; Moore, Eugene F; Savard, Guy

    2005-01-01

    An upgrade to the radioactive beam capability of the ATLAS facility has been proposed using 252Cf fission fragments thermalized and collected into a low-energy particle beam using a helium gas catcher. In order to reaccelerate these beams an existing ATLAS ECR ion source will be reconfigured as a charge breeder source. A 1Ci 252Cf source is expected to provide sufficient yield to deliver beams of up to ~106 far from stability ions per second on target. A facility description, the expected performance and the expected performance will be presented in this paper. This work is supported by the U.S. Department of Energy, Office of Nuclear Physics, under contract W-31-109-ENG-38.

  20. LARA. Localization of an automatized refueling machine by acoustical sounding in breeder reactors - implementation of artificial intelligence techniques

    International Nuclear Information System (INIS)

    The automatic control of the machine which handles the nuclear subassemblies in fast neutron reactors requires autonomous perception and decision tools. An acoustical device allows the machine to position in the work area. Artificial intelligence techniques are implemented to interpret the data: pattern recognition, scene analysis. The localization process is managed by an expert system. 6 refs.; 8 figs

  1. Preliminary design of core plasma parameters for the fusion-fission hybrid reactor based on GDT

    International Nuclear Information System (INIS)

    Based on the recent experiment progress of Gas Dynamic Trap (GDT), a core plasma physics conceptual design for driving fission blanket was proposed. The 0-D physical model was built and the core plasma parameters with 50 MW fusion power were preliminarily designed. The reliability of the physical model and design was demonstrated by comparison between the calculation and the experiment results. (authors)

  2. Methods and devices prepared to eliminate activation and fission products from PEC reactor cover gas

    International Nuclear Information System (INIS)

    The major effort made in Italy for the development of fast nuclear reactor is concentrated in the PEC reactor, whose construction is now in the completion stage. The PEC reactor (Prova Elementi di Combustibile - Fuel Element Testing ) is a sodium-cooled reactor with a power rating of 120 MWt, being built for the purpose of studying the behavior of fuel elements under thermal and neutronic conditions similar to those of fast reactor power stations, whit particular attention to safety aspects. The PEC reactor represents a research instrument particularly suitable for studies and experiments in the following fields: performances of the fuel element and its economical optimization (also with the possibility of testing fuel elements not necessarily based on mixed oxides); experiments in the safety field, not only referred to fuel elements, but also to plant subsystems. The experimental program will cover the research of the limit conditions of the typical parameters, such as cladding temperature, linear power, radiation rate, etc. PEC will also allow researches on new-concept fuel elements and thermal, hydraulic and power transients and cycles foreseen in the commercial power plants under normal, upset and emergency conditions. A number of the solutions regarding the PEC reactor and preparatory approaches to its operation are reported in this paper. In particular the following items are discussed: a description of three cover-gas circuits present in the reactor; an estimate of the contamination conditions foreseen under operating conditions; a description of the equipment for the purification of the cover gas and relative operating conditions. There are three cover-gas circuits present in the PEC reactor. They concern the following sodium circuits: primary reactor, primary emergency reactor and sodium purification primary reactor; secondary reactor, test channel and emergency reactor; primary test channel

  3. Gamma dose mapping fuelling machine at Dhruva reactor using ESR dosimetry

    International Nuclear Information System (INIS)

    Electron Spin Resonance (ESR) technique is useful in quantifying the radiation induced radicals with unpaired electrons, as a measure of irradiation dose delivered, thereby has application in the field of dosimetry. The ESR dosimetry has a few reported dosimeters, Alanine, Lithium Carbonate are commonly used to elucidate dosimetric response by the ESR spectrometer. We have identified, reported the ESR dosimetric properties of glass powder and hence, glass powder was used along with Alanine to calibrate its response and for dose mapping studies. The fuelling machine at Dhruva reactor during the hot rod handling has points that exhibit high radiation field thereby conventional dosimetric study is not feasible. Since ESR dosimetry has viability for high dose ranges, it was decided to carry out gamma dose mapping of select areas of the fuelling machine facility. The gamma dose mapping results at various places of fuelling machine of the reactor was estimated, Since the irradiation time for the either set of samples is known, the radiation fields could be worked out at the various points as 40-60 R/h, and the dosimetric response of alanine and glass powder samples were found to tally within ± 10%. ESR dosimetry has been useful in the dose mapping of high radiation areas, and the new ESR dosimetric material-glass powder, has nearly the same dose response as that of alanine

  4. Man-machine systems research at the OECD Halden reactor project

    International Nuclear Information System (INIS)

    The OECD Halden Reactor Project is a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency with fifteen participating countries. One of the main research topics focuses on man-machine systems. Particular attention is paid to the operator's tasks in the reactor control room environment. The overall objective of the research in this field is to provide a basis for improving today's control rooms through the introduction of computer-based solutions for the effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. The programme comprises four main activities: the verification and validation of safety critical software systems; man-machine interaction research emphasizing improvements in man-machine interfaces on the basis of human factors studies; computerised operator support systems assisting the operator in fault detection/diagnosis and planning of control actions; and control room development providing a basis for retrofitting of existing control rooms and for the design of advanced concepts

  5. Optimal design study of cylindrical finned reactor for solar adsorption cooling machine working with activated carbon-ammonia pair

    International Nuclear Information System (INIS)

    This paper presents a model describing the heat and mass transfer in cylindrical finned reactor of solar adsorption refrigerator. Giving the meteorological data as boundary conditions on the reactor; the model computes the solar coefficient of performance (COPs). The validity of the model has been tested by using experimental results. An analysis of the sensitivity of the COPs versus the geometrical parameters of the reactor (radius of the reactor, fins thickness and fins number) is mad. Then the model is applied to optimize the solar reactor. The COPs is used as an optimization criterion. The geometrical parameters where the COPs of the machine reach a maximum have been calculated

  6. Energy released in fission

    International Nuclear Information System (INIS)

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

  7. The fission power of a conceptual fluidised bed thermal nuclear reactor

    International Nuclear Information System (INIS)

    The fluidised bed thermal nuclear reactor investigated in this paper is an innovative reactor design in which 1 mm diameter TRISO-coated fuel particles are fluidised by helium gas coolant in a 2,5 m diameter and 6 m high cylindrical bed. The coolant flow rate provides part of the reactivity control mechanism. The TRISO-coated particles have an enriched uranium oxide kernel surrounded by layers of porous carbon, pyrolytic carbon and silicon carbide. This paper presents detailed transient modelling results of this conceptual fluidised bed thermal nuclear reactor obtained using the FETCH nuclear criticality model. Previous work has provided evidence to suggest that such a reactor can be dynamically stable for low power outputs of ∝20 MWt. This work focuses on a reactor with a much higher thermal output of 100 MWt. To simulate the fluidised bed reactor the FETCH model has been used to solve the neutron transport equation in full-phase space, coupled to multi-phase gas-particle fluid dynamics. The main difficulty in modelling such a reactor is that its reactivity is a sensitive function of the fuel particle distribution inside the inner fluidised bed reactor cavity. This fuel particle distribution varies chaotically with time which is the root cause of the reactor's power variability. (orig.)

  8. Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

    Science.gov (United States)

    Clement, J. D.

    1973-01-01

    Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

  9. Regulation of the fission product activity in the primary coolant and assessment of defective fuel rod characteristics in steady-state WWER-type reactor operation

    International Nuclear Information System (INIS)

    Regulation of the maximum limiting levels of fuel cladding failure and normalizing of the fission product activity in the primary coolant of WWER-type reactors for steady-state reactor operation is considered. It is shown that for the advanced nuclear power plants with WWER-type reactors the maximum permissible level of fuel rod failure and fission product activity in the primary coolant must be determined taking into account the actual level of the fuel rod reliability, possible failures of other reactor equipment (steam generator tubes) and efficiency of the primary and secondary coolant purification systems. The computer code TIMS developed in RRC 'Kurchatov Institute' for the assessment of the number of failed fuel rods and defect characteristics by comparing measured and calculated values of fission product activities in primary coolant is described. The important feature of the code is the increase of reliability of assessment by taking into account the actual errors of fission product activity measurements and possible contamination of primary circuit. (author)

  10. Irradiated uranium reprocessing, Final report I-VI, IV Deo IV - Separation of uranium, plutonium and fission products from the irradiated fuel of the reactor in Vinca

    International Nuclear Information System (INIS)

    This study describes the technology for separation of uranium, plutonium and fission products from the radioactive water solution which is obtained by dissolving the spent uranium fuel from the reactor in Vinca. The procedure should be completed in a hot cell, with the maximum permitted activity of 10 Ci

  11. A loading machine for fuel assemblies in nuclear reactor core, equipped with removable guiding and travelling means

    International Nuclear Information System (INIS)

    The charging machine is composed of a guiding and displacement set for the internal mobile mast which comprises a fixed tubular drum laying on the loading machine carriage, a tubular rotating drum placed inside the fixed drum, vertically guiding means for the mobile mast and a support for the lifting means, jointly rotating with the rotating drum

  12. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  13. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%

  14. Studies on short-lived fission products at the Mainz TRIGA reactor

    International Nuclear Information System (INIS)

    Neutron-rich nuclei of medium mass number are produced by thermal-neutron-induced fission of heavy elements, e.g., 235U, 239Pu, and 249Cf. Pulse irradiations lead to an enhancement of the ratio of short-lived activities to the accompanying longer-lived components. One approach for investigating the properties of short-lived nuclei consists in a combination of rapid chemical separations with higher-resolution gamma spectroscopy. This is demonstrated by the isolation of neutron-rich isotopes of niobium by sorption on glass and of ruthenium by solvent extraction. Other rapid separation procedures from aqueous solutions are briefly summarized and a few examples for their application in nuclear fission- and delayed neutron studies are given. Some experiments with an on-line mass separator of the ISOLDE-type, using chemical targets, are described. (U.S.)

  15. Separation of gaseous fission products from reactor and reprocessing-plant off-gases

    International Nuclear Information System (INIS)

    Considerable amounts of gaseous radioactive fission products krypton and xenon result from nuclear power generation, thus creating a radiation hazard in the event of their uncontrolled release to the environment. The most efficient method for treating off-gas streams appears to be low-temperature distillation, the efficiency of which depends on an exact knowledge of vapor-liquid equilibrium data. Isothermal data are measured for mixtures of Kr and Xe between 145 and 170 K and for their mixtures with components of air between 100 and 125 K, which make possible the exact design of a distillation column suitable for effective removal of the fission products. The solubilities of Kr in the components of air and of Xe in liquid krypton are studied, and possible applications for the safe separation and fixation of Kr are discussed

  16. Measurement of airborne fission products in Chapel Hill, NC, USA from the Fukushima I reactor accident

    CERN Document Server

    MacMullin, S; Green, M P; Henning, R; Holmes, R; Vorren, K; Wilkerson, J F

    2011-01-01

    We present measurements of airborne fission products in Chapel Hill, NC, USA, from 62 days following the March 11, 2011, accident at the Fukushima I Nuclear Power Plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products I-131 and Cs-137 were measured with maximum activities of 4.2 +/- 0.6 mBq/m^2 and 0.42 +/- 0.07 mBq/m^2 respectively. Additional activity from I-131, I-132, Cs-134, Cs-136, Cs-137 and Te-132 were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  17. Potential for the use of hydrochloric acid in fission reactor fuel recycle

    International Nuclear Information System (INIS)

    The chemistry and the effects of the use of hydrochloric acid as the aqueous phase in fuel recycle are surveyed. Available data are sufficient to suggest that separations of actinides and fission products can be at least as good in an HCl-trialkyl amine system as in Purex. Advantages of the HCl system are simpler operations of the off-gas system, better separation of neptunium from uranium and plutonium, better control of oxidation states of the dissolved species, and simpler recycle of the acid. A possible advantage is the more complete dissolution of the fission products, leaving very little insoluble residue. Disadvantages include lack of development of methods for dissolution of oxide fuel in hydrochloric acid, the requirement for processing equipment constructed of titanium, possible complications in the waste-handling system, and the dissolution of much of the cladding in the case of stainless-steel clad fuel

  18. Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

    International Nuclear Information System (INIS)

    This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

  19. On-Line Fission Gas Release Monitoring System in the High Flux Reactor Petten

    International Nuclear Information System (INIS)

    For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed dubbed the 'Sweep Loop Facility' (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experimental rigs. The SLF enables continuous and independent surveillance of all gas circuits. The release of volatile fission products (FP) from the in-pile experiments is monitored by continuous gas purging. The fractional release of these FP, defined as the ratio between release rate of a gaseous fission isotope (measured) to its instantaneous birth rate (calculated), is a licensing-relevant test for HTR fuel. The developed gamma spectrometry station allows for higher measurement frequencies, thus enabling follow-up of rapid and massive release transients. The designed stand-alone system was tested and fully used through the final irradiation period of the HFR-EU1 experiment which was terminated on 18 February 2010. Its robustness allowed the set up to be used as extra safety instrumentation. This paper describes the gas activity measurement technique based on HPGe gamma spectrometry and illustrates how qualitative and quantitative analysis of volatile FP can be performed on-line. (authors)

  20. Energy storage and transfer with homopolar machine for a linear theta-pinch hybrid reactor

    International Nuclear Information System (INIS)

    This report describes the energy storage and transfer system for the compression coil system of a linear theta-pinch hybrid reactor (LTPHR). High efficiency and low cost are the principal requirements for the energy storage and transfer of 25 MJ/m or 25 GJ for a 1-km LTPHR. The circuit efficiency must be approximately 90 percent, and the cost for the circuit 5 to 6 cents/J. Scaling laws and simple relationships between circuit efficiency and cost per unit energy as a function of the half cycle time are presented. Capacitors and homopolor machines are considered as energy storage elements with both functioning basically as capacitors. The advantage of the homopolar machine in this application is its relatively low cost, whereas that of capacitors is better efficiency

  1. Investigations on the gettering of metallic fission products in the primary circuit of a high temperature reactor

    International Nuclear Information System (INIS)

    A new concept of gettering Ag-110 m and Cs-134 137 in the primary circuit of a High Temperature Reactor (HTR) is presented. It is based upon the known fact that the vapor pressure of metallic fission products in solid or liquid solutions is lower compared with that of the pure fission products. Although metallic additives were found not to influence the silver release from oxide fuel kernels, the effective diffusion coefficient of Ag-110 m in graphite matrix is reduced by about two orders of magnitude by small amounts of the metallic Cu, Ge, Sn or Au additions. However, these reduced silver diffusion coefficients are not sufficiently low in order to retain Ag-110 m in the fuel-free zone of spherical HTR fuel elements. On the other hand, metallic additives were found to be very efficient in gettering Ag-110 m from the gaseous pahse: During a contact time of only 0.15 seconds at 950sup(o)C more than 80%, at 850sup(o)C even more than 99% of the Ag-110 m could be absorbed from the streaming gas by using a metal-containing graphite filter. The best results were obtained by using Sn or Au additives. By optimizing the filter geometry further increase of the efficiency should be possible. (orig./HP)

  2. Air sampling system for evaluating the thyroid dose commitment due to fission products released from reactor containment. Final report

    International Nuclear Information System (INIS)

    Accidental releases of radioactivity from fission reactors will consist of active vapors and aerosols. Composition of the released plume or cloud will depend on the energy of release and fission product volatility. In accidents at Windscale and SL-1, 131I was the predominant isotope present in both the initial cloud and later release. Thus an air sampling system was developed for efficient radioiodine collection. The air sampling, readout, and dose assessment system was developed to be used in the environment after loss of containment accidents. The system can detect less than 1 rem dose commitments to thyroids of 5 year old children for immersion times of 10 hours or less. The air mover can be operated on either 110V ac power or 12V dc power available from vehicles with cigar lighter sockets. An inorganic silver loaded silica gel adsorber was developed for high mehyl iodine, HOI, and elemental iodine efficiency and low noble gas efficiency. A peal away high efficiency particulate filter permits the gaseous and particulate sample fractions to be evaluated separately. Predicted particulate iodine is combined with the adsorbed component to account for the total radioiodine in a given sample

  3. Multi-scale computer simulation of fission gas discharge during loss-of-flow accident in sodium fast reactor

    International Nuclear Information System (INIS)

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present recent results of computer simulations using a newly developed multi-scale multidimensional model of fission gas discharge following cladding failure during a postulated loss-of-flow accident in a Gen. IV Sodium Fast Reactor (SFR). The issues discussed in the full paper will include an overview of the proposed multi-scale three-dimensional (3D) modeling approach to the inter-related phenomena of transient fuel element heat-up, cladding failure mechanisms, injection of a jet of gaseous fission products into a partially blocked SFR coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using three different inter-communicating computational multiphase fluid dynamics (CMFD) codes: FronTier, PHASTA and NPHASE-CMFD. FronTier is a multi-physics code for the simulation of multiphase/free-surface flows based on the method of front tracking, which has been developed at SUNY at Stony Brook in collaboration with BNL and LANL. PHASTA is a parallel, hierarchic (between 2.- and 5. orders of accuracy, depending on function choice), adaptive, stabilized (finite element) transient analysis DNS flow solver (both incompressible and compressible) that has been developed at RPI. The PHASTA code uses anisotropic adaptive algorithms and the most advanced LES/DES models. NPHASE-CMFD is a robust computational multiphase fluid dynamics flow solver. The technology used by the NPHASE-CMFD code is the multi-field model of multiphase flows. The governing equations of fluid flow and heat transfer are ensemble-averaged, which allows the NPHASE-CMFD code to predict local non

  4. Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor.

    Science.gov (United States)

    Liu, H B; Brugger, R M; Rorer, D C; Tichler, P R; Hu, J P

    1994-10-01

    Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed. PMID:7869995

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

    International Nuclear Information System (INIS)

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

  6. Determination of burnup balance for nuclear reactor fuel on the basis of γ-spectrometric determination of fission products

    International Nuclear Information System (INIS)

    Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data

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

    International Nuclear Information System (INIS)

    Neutronics calculations were performed to analysis the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS-EM (Energy Multiplier) blanket. The most significant and main goal of the water-cooled FDS-EM blanket is to achieve the energy gain of about 1 GW with self-sustaining tritium, which can operate for as long as possible without fuel unloading and reloading. The preliminarily designed neutronics parameters for FDS-EM were presented, which show that the blanket loaded with the Nuclear Waste (transuranic from 33 000 MWD/MTU PWR and depleted uranium) for energy multiplication (M≅90) with tritium self-sufficiency can operate for at least 10 years without fuel unloading and reloading. (authors)

  8. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  9. Fission production and actinides in the spent graphite of the reactor stacks of the Siberian chemical integrated plant

    International Nuclear Information System (INIS)

    The peculiarity of the accomplished studies consisted in the representative selection of the reactor graphite stacks samples and in the performance of the complex analysis of their radioactive contamination. The role of incidents in forming the graphite contamination by individual radionuclides is identified and their distribution in stacks is studied. The correlation between the content of various radionuclides is investigated. The schemes for evaluating their reserve in the graphite stack are plotted. The results on evaluating the radionuclides reserve in the graphite stack highly differ from the earlier forecasted ones. The fission products and actinides reserves are by 10 times lesser as it was fore coated earlier, which may essentially simplify dismantling and selection of utilization technologies

  10. Renovation of new fuel transfer machine in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    In the higher performance plan (MK-III plan) of the experimental Fast Reactor JOYO, fuel handling system has been renovated to remote control system to reduce refueling time. As a part of this plan, new fuel transfer machine which is used to receive and transport new fuel, has been renovated completely to remote an automatic control system with no local operation and no local watching by Kawasaki Heavy Industries, Ltd. In this paper, the design and fabrication of this system are described. (author)

  11. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  12. A model for predicting fission product activities in reactor coolant: application of model for estimating I-129 levels in radioactive waste

    International Nuclear Information System (INIS)

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor. The latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of short-lived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analysed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-129 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines. This assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137, which are consistent with values reported for pressurised water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was10-8 - 10-7. (author)

  13. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    International Nuclear Information System (INIS)

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7

  14. Results with the electron cyclotron resonance charge breeder for the 252Cf fission source project (Californium Rare Ion Breeder Upgrade) at Argonne Tandem Linac Accelerator System

    International Nuclear Information System (INIS)

    The construction of the Californium Rare Ion Breeder Upgrade, a new radioactive beam facility for the Argonne Tandem Linac Accelerator System (ATLAS), is nearing completion. The facility will use fission fragments from a 1 Ci 252Cf source; thermalized and collected into a low-energy particle beam by a helium gas catcher. In order to reaccelerate these beams, an existing ATLAS electron cyclotron resonance (ECR) ion source was redesigned to function as an ECR charge breeder. Thus far, the charge breeder has been tested with stable beams of rubidium and cesium achieving charge breeding efficiencies of 9.7% into 85Rb17+ and 2.9% into 133Cs20+.

  15. Delayed neutron measurements of induced fission rates in burnt LWR fuel samples at the Proteus zero-power reactor facility - 125

    International Nuclear Information System (INIS)

    The LIFE'at'PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel, following re-irradiation in the zero-power PROTEUS research reactor. In the presented approach, the fission rates are estimated by measuring delayed neutrons emitted by re-irradiated fuel. To demonstrate the feasibility of this technique, fresh and burnt fuel samples (with burnup varying from 36 to 64 GWd/MTU) were irradiated in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Relative fission rates between different core lattice positions were derived for a fresh sample as well as for the three burnt samples. The measured fission rate ratios have 1-σ uncertainties between 2% and 3.5%, with the larger uncertainties corresponding to the more highly burnt fuel. Results obtained by Monte Carlo simulations agree with the experimentally determined values within these limits. With further development of the technique, the experimental uncertainties can be further reduced. Continuing effort is being directed towards accurate comparison of fission rates between fuel samples of different burn-up. (authors)

  16. Numerical analysis of the release of metallic fission product in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    HTGR core should be designed under the limit of circulating coolant activity and plateout activity in the primary circuit. Release of metallic fission product which is important in plateout activity strongly depends on a geometry of coated particle fuel and on a space distribution and time history of temperature, etc. Therefore, many iterations of a series of nuclear, thermal and fission product release analyses are required for selection of suitable fuel and core dimensions. At the stage of conceptual core design, however, it seems to be impossible to execute these enormous iterative analyses. In this paper, for the purpose of alleviating the troublesome routines, a simplified calculating procedure of Cs137 release, the most important metallic fission product, is presented as the results of the parametric survey analogues to sensitivity analysis. At first, several parameters considered to play a fair role in Cs137 release are selected and are varied independently in sensitivity analysis. Secondly, on the bases of analytic results we pick out the main parameters that have significant effects on Cs137 release. The results show: (i) the characteristics of Cs137 release can be explained by the time-averaged core maximum temperature: T-tildefmax; (ii) the dominant release is that from 'intact particles' rather than tailed particles in the range of T-tildefmax>1300 deg. C under the assumption of failure fraction 0.1%; (iii) failure fraction up to 2% has no significant effect on Cs137 release at T-tildefmax=1400 deg. C; (iv) Cs137 release decreases to 1/5 when the thickness of SiC layer increases from 25μ to 50μ; (v) in comparison with pin-in-block fuel type, multi hole fuel type is lower in Cs137 release. (author)

  17. Fission - track age of the Marjalahti Pallasite

    International Nuclear Information System (INIS)

    Full text: Investigation of fossil charged-particle tracks in various mineral phases of extraterrestrial samples is a powerful method for research the early stages of the solar system. Over geological time, meteorites crystals have accumulated a record of tracks produced by heavily charged energetic particles from both internal (spontaneous fission of 238U and some other extinct isotopes) and external sources (galactic cosmic rays with Z>20). The fortunate fact that meteorite grains can accumulate latent and very long-lived tracks since soon after the end of nucleosynthesis in the solar nebula enables one to decode their radiation history and to detect any thermal events in the meteorite cosmic history by revealing these tracks through suitable etching procedures. Only a few minerals in meteorites (mainly phosphates) contain small amount of uranium; the fact that 238U undergoes fission with fission-decay constant λf∼8.2x10-17 yr-1 allows one to use this isotope as a chronometer. By measuring the U concentration in the crystals (by reactor irradiation) and the density of the spontaneous-fission tracks it is relatively easy to calculate the 'fission-track age' if 238U is the main source of fission tracks. However the fission-track dating of extraterrestrial samples compared with the terrestrial ones has some peculiar features due to presence of a number of other potential track sources except the spontaneous fission of 238U, such as the spontaneous fission of presently extinct 244Pu, heavy nuclei of cosmic rays and induced fission by cosmic ray primaries. Only tracks from the spontaneous fission of U and Pu are suitable for fission-track dating. The competing effects of these fissioning elements, whose half-lives differ by a factor of ∼50, form a basis for a fission-track chronology for samples older than ∼ 4.0 Gyr. Over small intervals in time (∼ few x108 yr ) the track density from spontaneous fission of 238U is nearly constant. However, the contribution

  18. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. 12-month progress report, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

  19. Eugene P. Wigner’s Visionary Contributions to Generations-I through IV Fission Reactors

    OpenAIRE

    Carré Frank

    2014-01-01

    Among Europe’s greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechan...

  20. A revaluation of helium/dpa ratios for fast reactor and thermal reactor data in fission-fusion correlations

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Oliver, B.M.

    1996-10-01

    For many years it has been accepted that significant differences exist in the helium/dpa ratios produced in fast reactors and various proposed fusion energy devices. In general, the differences arise from the much larger rate of (n,{alpha}) threshold reactions occurring in fusion devices, reactions which occur for energies {ge} 6 MeV. It now appears, however, that for nickel-containing alloys in fast reactors the difference may not have been as large as was originally anticipated. In stainless steels that have a very long incubation period for swelling, for instance, the average helium concentration over the duration of the transient regime have been demonstrated in an earlier paper to be much larger in the FFTF out-of-core regions than first calculated. The helium/dpa ratios in some experiments conducted near the core edge or just outside of the FFTF core actually increase strongly throughout the irradiation, as {sup 59}Ni slowly forms by transmutation of {sup 58}Ni. This highly exothermic {sup 59}Ni(n,{alpha}) reaction occurs in all fast reactors, but is stronger in the softer spectra of oxide-fueled cores such as FFTF and weaker in the harder spectra of metal-fueled cores such as EBR-II. The formation of {sup 59}Ni also increases strongly in out-of-core unfueled regions where the reactor spectra softens with distance from the core.

  1. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    International Nuclear Information System (INIS)

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs

  2. Nuclear Fission Reactor Safety Research in FP7 and future perspectives

    CERN Document Server

    Garbil, Roger

    2014-01-01

    The European Union (ЕU) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

  3. Natural fission reactors in the Franceville basin, Gabon: A review of the conditions and results of a open-quotes critical eventclose quotes in a geologic system

    International Nuclear Information System (INIS)

    Natural nuclear fission reactors are only known in two uranium deposits in the world, the Oklo and Bangombe deposits of the Franceville basin: Gabon. Since 1982, five new reactor zones have been discovered in these deposits and studied since 1989 in a cooperative European program. New geological, mineralogical, and geochemical studies have been carried out in order to understand the behavior of the actinides and fission products which have been stored in a geological environment for more than 2.0 Ga years. The Franceville basin and the uranium deposits remained geologically stable over a long period of time. Therefore, the sites of Oklo and Bangombe are well preserved. For the reactors, two main periods of actinide and radionuclides migration have been observed: during the criticality, under P-T conditions of 300 bars and 400-500 degrees C, respectively, and during a distention event which affected the Franceville basin 800 to 900 Ma ago and which was responsible for the intrusion of dolerite dikes close to the reactors. New isotopic analyses on uranium dioxides, clays, and phosphates allow us to determine their respective importance for the retention of fission products. The UO2 matrix appears to be efficient at retaining most actinides and fission products such as REEs, Y, and Zr but not the volatile fission products (Cd, Cs, Xe, and Kr) nor Rb, Sr, and Ba. Some fissiogenic elements such as Mo, Tc, Ru, Rh, Pd, and Te could have formed metallic and oxide inclusion in the UO2 matrix which are similar to those observed in artificial spent fuel. Clays and phosphate minerals also appear to have played a role in the retention of fissiogenic REEs and also of Pu. 82 refs., 21 figs., 12 tabs

  4. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    International Nuclear Information System (INIS)

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical

  5. Prediction of nongaseous fission products behavior in the primary cooling system of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified. This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model. (author)

  6. Energy production using fission fragment rockets

    Science.gov (United States)

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

  7. Energy production using fission fragment rockets

    International Nuclear Information System (INIS)

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs

  8. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle

    International Nuclear Information System (INIS)

    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  9. Man-machine systems research at the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    The OECD Halden Reactor Project is a joint undertaking of national organisations in 15 countries sponsoring a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency. One main research area is man-machine systems addressing the operator tasks in the control room environment. The overall objective of these efforts is to provide a basis for improving today's control rooms through introduction of computer-based solutions for effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. The programme comprises four main activities: 1) verification and validation of safety critical software systems; 2) man-machine interaction research emphasizing improvements in man-machine interfaces on the basis of human factors studies; 3) computerized operator support systems assisting the operator in fault detection/diagnosis and planning of control actions; and 4) control room development providing a basis for retrofitting of existing control rooms and for the design of advanced concepts. (author). 10 refs, 3 figs

  10. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    Science.gov (United States)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  11. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    International Nuclear Information System (INIS)

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa–232U–233U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production

  12. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, A. N., E-mail: shmelan@mail.ru; Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Kurnaev, V. A., E-mail: kurnaev@yandex.ru; Salahutdinov, G. H., E-mail: saip07@mail.ru; Kulikov, E. G., E-mail: egkulikov@mephi.ru; Apse, V. A., E-mail: apseva@mail.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  13. Eugene P. Wigner’s Visionary Contributions to Generations-I through IV Fission Reactors

    Directory of Open Access Journals (Sweden)

    Carré Frank

    2014-01-01

    Full Text Available Among Europe’s greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

  14. Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

    International Nuclear Information System (INIS)

    Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric) hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of Ip = 15 MA, and toroidal field on the axis of Bo = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is ∼7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs

  15. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. Addendum 1. Alternate concepts. 12-month progress report addendum, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    During the course of the Mirror Hybrid Fusion-Fission Reactor study several alternate concepts were considered for various reactor components. Several of the alternate concepts do appear to exhibit features with potential advantage for use in the mirror hybrid reactor. These are described and should possibly be investigated further in the future

  16. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H.; Jiang, Weilin

    2014-11-01

    MAX phases, such as titanium silicon carbide (Ti3SiC2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti3SiC2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti3SiC2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti3SiC2, SiC, and a dual-phase nanocomposite of Ti3SiC2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti3SiC2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti3SiC2 occurs during ion implantation at 873 K. Cs in Ti3SiC2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti3SiC2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.

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

    DEFF Research Database (Denmark)

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

    1985-01-01

    Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...

  18. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  19. Preliminary results of the BTD-105B experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    International Nuclear Information System (INIS)

    The BTF-105B test is one of a series of in-reactor, all-effects, tests performed to measure fuel behaviour, fission-product release, and transport from nuclear fuel subjected to accident conditions. The primary objective of the BTF-105B test was to determine the timing, amount, and transport characteristics of fission products released from a previously irradiated CANDU-sized fuel element subjected to a high-temperature transient, representative of a loss-of-coolant accident with loss-of-emergency-core-cooling conditions, at an average fuel temperature of 1800 deg C. The preliminary results of the BTF-105B test are presented in this paper. The results include process parameters, from which boundary conditions could be derived for simulating the test; fuel parameters, such as sheath temperatures and hydrogen production rate; and fission-product release and transport data. (author)

  20. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    International Nuclear Information System (INIS)

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity

  1. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2003-10-31

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity.

  2. The integration of D-cycle tokamaks with decentralized small fusion and fission reactors

    International Nuclear Information System (INIS)

    A reaction physics analysis is undertaken of the conceptual characteristics of a large tokamak fusion reactor, equipped with a breeding blanket and operating on the deuterium cycle, supplying helium-3 fuel for numerous small D-3He fusion satellite reactors, as well as fissile fuel for selfregulating small nuclear heat sources. A range of systems parameters consistent with current experimental projections and design ranges are incorporated in the analysis. It is found that the intrinsic reaction and energetics complementarity provides a scientific basis for further development of this concept. The two dominant and particularly appealing features of this concept are the expanded latitude in system design and the prospects of introducing small decentralized nuclear energy sources for remote and special purpose applications. (orig.)

  3. On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

  4. Decrease of the CANDU spent nuclear waste inventories in fusion-fission (hybrid) reactors

    International Nuclear Information System (INIS)

    The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D,T) and catalyzed (D,D) modes. The analysis is conducted for a CANDU spent nuclear fuel which was used up to a total enrichment grade of 0.418%. The behavior of the spent fuel is observed during 48 months for discrete time intervals of Δt = 6 months. The cooling of the fissile fuel zone is considered with three different coolants, notably gas (preferably He), Flibe (Li2BeF4) and natural lithium. A rejuvenation period of 8 months is evaluated for a final fissile fuel enrichment grade of 1% for all coolant types in the fissile zone under a first-wall fusion neutron current load of 1,014 (2.45-MeV n/cm2.s) and 1,014 (14.1-MeV n/cm2.x), corresponding to 2.64 MW/m2 by a plant factor of 75% for the catalyzed (D,D) fusion reactor mode. The rejuvenation period increases to 12 months for the same fissile fuel enrichment grade using the (D,T) fusion reactor mode under a first-wall fusion neutron current load of 1,014 (2.45-MeV n/cm2.s), corresponding to 2.25 MW/m2 by a plant factor of 75%. This enrichment would be sufficient for a re-utilization in a CANDU reactor

  5. Calculation of fission product behavior in a multiple reactor barriers in case of an accident

    International Nuclear Information System (INIS)

    Radiation protection of the population in case of a reactor accident utilizes reference levels which are based on doses values. Therefore, adequate provisions for effective and timely dose assessment for population in case of accidents at nuclear power plant (NPP) are important. Developing the background for such provisions is the objective of this study. In particular, an exponential model has been developed and utilized to calculate the release rate of the most volatile gaseous materials from different reactor barriers. Calculation has been performed for noble gases (133Xe, 135Xe, 138Xe, 85Kr, 87Kr, 88Kr) and the halogens(1'31I, 132I, 133I, 1'34I, 135I). The effective dose rate equivalent is calculations in the nearly stage of a reactor accident. Calculations are performed using the MCNP-4C code. The results are comparable with the final analysis report which utilizes different codes. Results of our calculation shows no excessive dose in populated regions and it is recommended to use secondary containment barrier for highly reduction of the release rate to the environment. (Author)

  6. Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Durand -Smet, R.; Lourme, P. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    - Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

  7. Seismic analysis of new fuel assembly loading machine for China experimental fast reactor

    International Nuclear Information System (INIS)

    New fuel assembly loading machine of China Experiment Fast Reactor is a kind of kinetic equipment with very complex structure. Many of its motional and driving components can not be simulated exactly by finite element model (FEM). A simplified FEM analysis method was introduced in the paper, and the main frame of the equipment was simulated by a simplified FEM model. Response spectrum analysis method was used to obtain the acceleration response of the main components of the equipment under seismic condition. Theoretical analysis method was used to calculate the stresses of the main connecting bolts, and these bolts were evaluated based the regulations of nuclear codes to ensure the structure integrity of the equipment. (authors)

  8. Home brew technetium : clinical scale desktop plasma fusion neutron source to produce Tc99m as an alternative to industrial scale fission reactor sources

    International Nuclear Information System (INIS)

    Full text: Tc-99m (decay product of Mo-99) accounts for ∼ 90% of world's production of radiopharmaceuticals. Recent unexpected shutdowns of two fission reactors and routine maintenance closures .e created a global shortage of Tc-99m, hence the large global effort to find alternative sources. This project aims to design and produce a novel prototype Mo-99/Tc-99m source. An operational desktop neutron source is available at the University of Sydney, employing a deuterium fusion-plasma to create 2.45 MeV neutrons. These neutrons will be used to activate Mo-98 thin an activation vessel. In one embodiment, the activation vessel contains an aqueous slurry or gel containing Mo-98 which converts to 0-99 upon activation. The decay product Tc-99m could then be milked, similar to existing Tc-99m generators. Monte Carlo will be :ed to assess yield versus size and geometry for various vessel designs. The neutron source filled with deuterium operating at 250 W, produces 3 x 106 neutrons continuously. The neutron flux can be increased ∼ 100-fold if the fill gas is 50% tritium and by another ∼ 100-1000-fold by increasing the power. This is being designed for local use, perhaps on the scale f one or a few hospitals, so the yield would not need to be industrial ;ale as with fission reactor sources. This device is low cost <$300 K) compared with cyclotrons and fission reactors.

  9. Development of a calculation method for one dimensional kinetic analysis in fission reactors, with feedback effects

    International Nuclear Information System (INIS)

    The methodology used in the WIGLE3 computer code is studied. This methodology has been applied for the steady-state and transient solutions of the one-dimensional, two-group, diffusion equations in slab geometry, in axial type probelm analysis. It's also studied, based in a WIGLE3 computer code, reactor representative models, considering non-boiling heat transfer. A steady-state program for control rod bank position search- CITER 1D- has been developed. Some criticality research on the proposed system has been done using different control rod bank initial positions, time steps and convergence parameters. (E.G.)

  10. Applications: fission, nuclear reactors. Safety, its importance assertion and its implementation

    International Nuclear Information System (INIS)

    The various safety levels for nuclear reactors are reviewed: the first one is related to the quality of design, the second one to the assurance of quality during operation, the third one to the safety measures limiting the system and material failure consequences, the fourth one to the management of severe accidents, and the fifth one to the population safety measures. In parallel to these deterministic measures, probabilistic studies have been also undertaken in order to evaluate risks dealing with the core fusion, radioactive product release and effects on population. Safety is also concerned with material and electronic material ageing, software reliability, fire risks and diphasic thermohydraulics

  11. Security of supply for fission medical radio-isotopes based on optimal use of the test reactor network

    International Nuclear Information System (INIS)

    Nuclear Medicine relies to a large extend (80 % of the procedures) on radioisotopes produced by fission of uranium, on Mo99 /Tc99m for 28 million diagnoses made annually all over the world for tracking diseases in cancerology, cardiology, neurology ... and on I131 and Y90 for 3 million therapy procedures. The only four main producers (95 % of the world demand) are relying on 5 aging test reactors for irradiating HEU targets to be processed for extracting these short life isotopes before their conditioning as radiopharmaceuticals to be daily used in hospitals. Ensuring the security of supply has been a challenge for many years and if several shortages occurred in the past, the last crises in 2007 and 2008 revealed more than ever the weakness of the current situation despite the efforts and warning that have been devoted to facing many obstacles including possible technical failures, incidents, transport constraints and licensing issues, as well as political threat for the use of HEU. It is time for having all stakeholders drawing the lessons of the crisis and considering all possible serious and realistic improvements on technical and organisational issues without neglecting the resulting economical and safety aspects. (author)

  12. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables

  13. Economics analysis of fuel cycle cost of fusion–fission hybrid reactors based on different fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2015-01-15

    Highlights: • Economics analysis of fuel cycle cost of FFHRs is carried out. • The mass flows of different fuel cycle strategies are established based on the equilibrium fuel cycle model. • The levelized fuel cycle costs of different fuel cycle strategies are calculated, and compared with current once-through fuel cycle. - Abstract: The economics analysis of fuel cycle cost of fusion–fission hybrid reactors has been performed to compare four fuel cycle strategies: light water cooled blanket burning natural uranium (Strategy A) or spent nuclear fuel (Strategy B), sodium cooled blanket burning transuranics (Strategy C) or minor actinides (Strategy D). The levelized fuel cycle costs (LFCC) which does not include the capital cost, operation and maintenance cost have been calculated based on the equilibrium mass flows. The current once-through (OT) cycle strategy has also been analyzed to serve as the reference fuel cycle for comparisons. It is found that Strategy A and Strategy B have lower LFCCs than OT cycle; although the LFCC of Strategy C is higher than that of OT cycle when the uranium price is at its nominal value, it would become comparable to that of OT cycle when the uranium price reaches its historical peak value level; Strategy D shows the highest LFCC, because it needs to reprocess huge mass of spent nuclear fuel; LFCC is sensitive to the discharge burnup of the nuclear fuel.

  14. Fission fragment rocket concept

    International Nuclear Information System (INIS)

    A new propulsion scheme is outlined which may permit interstellar missions for spacecraft. This scheme is based on the idea of allowing fission fragments to escape from the core of a nuclear reactor. (orig.)

  15. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  16. Local fission gas release from high burnup water reactor fuel under transient conditions

    International Nuclear Information System (INIS)

    The paper presents results for local gas release, produced by power transients (bump tests) at the end of life of a water reactor fuel. The burnup was from 2.7-4.5% FIMA (25,000-41,000 MWd/tU). The local linear power at the bump terminal level (BTL) of the fuel examined ranged from 300-415 W/cm. The hold time at BTL was either 24 or 72 h. Around 410-415 W/cm, the local gas releases measured on pellet sized samples were 35-40%. Radial xenon release profiles measured by electron microprobe analysis showed that onset of release occurred at about 700 deg. C. Above 1100 deg. C, a constant release of about 95% was found. (author)

  17. Measurement of fission product activity in the Peach Bottom Reactor primary coolant loop

    International Nuclear Information System (INIS)

    The distribution of gamma-emitting radionuclides deposited in the primary circuit of the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) at end-of-life has been determined by in situ gamma scanning. The work was part of the Peach Bottom End-of-Life Program and was performed by the IRT Corporation under subcontract to General Atomic Company. The measurements were made to support a design method verification exercise. The specific activity on the ducts was measured by external scans at local points with a Ge(Li) detector and by internal scans with a travelling intrinsic germanium detector (after destructive removal of trepan samples); the activity on the steam generator tube bundle was determined by traversing selected tubes with travelling CdTe detectors from the water side. Calibration measurements on mockups allowed reduction of the spectra to specific activity

  18. Fission energy program of the US Department of Energy, FY 1981

    International Nuclear Information System (INIS)

    Information is presented concerning the National Energy Plan and fission energy policy; fission energy program management; converter reactor systems; breeder reactor systems; and special nuclear evaluations and systems

  19. Fission energy program of the US Department of Energy, FY 1981

    Energy Technology Data Exchange (ETDEWEB)

    Ferguson, Robert L.

    1980-03-01

    Information is presented concerning the National Energy Plan and fission energy policy; fission energy program management; converter reactor systems; breeder reactor systems; and special nuclear evaluations and systems.

  20. Status of fission yield measurements

    International Nuclear Information System (INIS)

    Fission yield measurement and yield compilation activities in the major laboratories of the world are reviewed. In addition to a general review of the effort of each laboratory, a brief summary of yield measurement activities by fissioning nuclide is presented. A new fast reactor fission yield measurement program being conducted in the US is described

  1. A kinetic model for fission-product release and fuel oxidation behaviour for zircaloy-clad fuel elements under reactor accident conditions

    International Nuclear Information System (INIS)

    During a severe reactor accident fission products will be released from the degraded fuel in the reactor core. In addition, hydrogen will be generated at high-temperature by the steam oxidation of the core materials. This oxidation process will also influence the rate of fission-product release. Separate-effects tests performed out-of-pile at the Chalk River Laboratories (CRL) have provided a better understanding of the processes of fission product release during severe accident conditions. The annealing experiments were conducted in steam at temperatures ranging from 1200 to 1700 deg C with irradiated fuel specimens of uranium dioxide in the form of bare fuel fragments and a short-length Zircaloy-clad fuel element. The fission product release was monitored by online gamma ray spectrometry. The oxygen partial pressure was also measured with solid-state oxygen sensors, providing a calculation of the rate of oxygen consumption and hydrogen production in the fuel specimens. Based on the CRL tests, an analytical model has been developed to describe the kinetic release behaviour of the volatile fission-product species (cesium) during high-temperature accident conditions. The physically-based model accounts for the kinetics of fuel oxidation as a rate-determining reaction at the fuel/steam interface. A more general framework is therefore provided to detail the influence of the atmosphere (i.e. oxygen potential) on the behaviour of the fission product release. Solid state diffusion in the fuel matrix is shown to be the rate-controlling mechanism in the early stages of release. The enhanced diffusivity of fission products in the hyperstoichiometric fuel is modelled with the assumption that diffusion takes place on vacant cation lattice sites. When the fuel reaches a state of oxidation of x ∼ 0.07 for the UO2+x phase, a more rapid release process occurs in accordance with first-order rate kinetics. The retarding influence of the hydrogen production on the fuel oxidation

  2. Oklo natural fission reactor program. Progress report, April 1-August 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, D.B. (comp.)

    1980-12-01

    An interim report has been published on the redistribution of uranium, thorium, and lead in samples representing several million cubic meters of sandstone and metamorphosed sediments in the Athabasca Basin which is located in the northwest corner of the Canadian province of Saskatchewan. The region of study includes zones of uranium mineralization at Key Lake. Mineralization occurs at the unconformity between the Athabasca sandstone and the underlying metasediments and in fault zones within the metasediments. Lead isotopes record a radiometric age of 1300 +- 150 m.y. in samples from above and below the unconformity. This age probably reflects the time of deposition of the sandstones and an associated redistribution of uranium and/or lead in the underlying rocks. Many of the samples have been fractionated with respect to radiogenic lead and the actinide parent elements since that time. Sandstones and altered rocks from the region above the unconformity have been a transport path and are a repository for lead. In contrast, mineralized rocks are deficient in radiogenic lead and must be an important source of lead in the local geologic environment. Samples from Oklo reactor zone 9 and nearby host rocks have been prepared for isotopic analyses of ruthenium, molybdenum, uranium and lead.

  3. Calculational study on neutron kinetics and thermodynamics of a gaseous core fission reactor. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1992-01-01

    The aim of the authors' work was to investigate the static and dynamic properties of a GCFR with oscillating (moving) fuel gas. A simplified schematic diagram of such a GCFR, similar to the concept of Kistemaker (Kis78a), is shown. It consists of a graphite cylinder of, say, 2 m diameter and 10 m length, filled with a mixture of uranium and carbon fluorides (UCF) at high temperature in ionized state, in chemical and thermodynamical equilibrium with the graphite cylinder wall (Kis78a, Kis86, Kle87). The cylindrical gas space is divided into an active 'core' region, surrounded by an effective (thick) neutron reflector, and a so-called 'expander' region, surrounded by a much less effective (thinner or with neutron poison) neutron reflector. In operation, part of the fuel gas oscillates back and forth between core and expander region. The investigation requires the study of neutron statics, neutron kinetics, reactor gas thermodynamics and gas dynamics, resulting in a combined calculational model, containing these aspects. In order to achieve this the authors followed a step-by-step approach.

  4. Retention of activation and fission radionuclides by mallards from the Test Reactor Area radioactive leaching pond

    International Nuclear Information System (INIS)

    Twenty semi-wild mallard ducks were banded, fitted with dorsal and ventral thermoluminescent dosimeter packets, and released on the Test Reactor Area radioactive leaching ponds. Ducks were live captured after 75 days and 145 days on the pond, placed in metabolic cages and whole-body counted periodically for 52 days. Ducks from each group were sacrificed immediately after capture, dissected, and muscle, feather, gut, and liver samples submitted for analyses. The remaining ducks were also sacrificed and dissected after the 52 day counting period. Concentrations of the 17 gamma emitting radionuclides detected at capture and after 52 days of physical and biological decay were compared. Highest mean radionuclide concentrations were found in feathers followed by gut, liver, and muscle. Effective and biological halflives of Zn-65, Cr-51, Cs-134, Cs-137, and Se-75 were determined and compared with data from previous studies. Samples are currently being analyzed for Pu-238, Pu-239-240, Am-241, Cm-242, Cm-244 and Sr-90. Further data analyses will be completed after data collection has terminated

  5. Technologies for tritium control in fission reactors moderated with heavy water

    International Nuclear Information System (INIS)

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  6. Tensile and electrical properties of copper alloys irradiated in a fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [Scientific Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States)] [and others

    1996-04-01

    Postirradiation electrical sensitivity and tensile measurements have been completed on pure copper and copper alloy sheet tensile specimens irradiated in the SM-2 reactor to doses of {approx}0.5 to 5 dpa and temperatures between {approx}80 and 400{degrees}C. Considerable radiation hardening and accompanying embrittlement was observed in all of the specimens at irradiation temperature below 200{degrees}C. The radiation-induced electrical conductivity degradation consisted of two main components: solid transmutation effects and radiation damage (defect cluster and particle dissolution) effects. The radiation damage component was nearly constant for the doses in this study, with a value of {approx}1.2n{Omega}m for pure copper and {approx}1.6n{Omega}m for dispersion strengthened copper irradiated at {approx}100{degrees}C. The solid transmutation component was proportional to the thermal neutron flux, and became larger than the radiation damage component for fluences larger than {approx}5 10{sup 24} n.m{sup 2}. The radiation hardening and electrical conductivity degradation decreased with increasing irradiation temperature, and became negligible for temperatures above {approx}300{degrees}C.

  7. The experimental determination and evaluation of the three-dimensional fission density distribution of the IPEN/MB-01 research reactor facility for the IRPhE project

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos, E-mail: asantos@ipen.br [Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, SP 05508-000 (Brazil); Fanaro, Leda C.C.B., E-mail: lcfanaro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, SP 05508-000 (Brazil); Andrade e Silva, Graciete S. de, E-mail: gsasilva@ipen.br [Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, SP 05508-000 (Brazil); Mendonca, Arlindo G., E-mail: amendon@ipen.br [Centro Tecnologico da Marinha em Sao Paulo, CTMSP, Av. Prof. Lineu Prestes, 2468, Cidade Universitaria, SP 05508-000 (Brazil)

    2011-02-15

    Experiments for the determination of the three-dimensional fission density distribution were successfully carried out at the IPEN/MB-01 research reactor facility. The experiment is of very good quality for utilization in the benchmark of computer codes and related nuclear data library commonly used for the calculation of fission density distribution in reactor cores. The complete experimental data set comprises a very massive set of measured fission density distribution. The experiment was evaluated and included in the IRPhE handbook. Two calculation methodologies were employed for the theoretical analysis of the proposed benchmark; one stochastic (MCNP-5) and the other one deterministic (NJOY/AMPX-II/TORT); both utilizing ENDF/B-VII.0 as the source of nuclear data. The theory/experiment comparison reveals in general a good agreement for both methodologies. However, MCNP-5 results are in a better shape than those of TORT. TORT underestimates the relative power distribution mainly in the axial upper part of the fuel rod. This effect is mostly credited to the over-prediction of the control bank worth.

  8. Volatile and gaseous fission-product source term evaluation during power ramping conditions in water reactor fuel using the mechanistic FASTGRASS computer code

    International Nuclear Information System (INIS)

    As the noble gases play a major role in establishing the interconnection of escape routes from the interior to the exterior of nuclear reactor fuel, a realistic description of the release of volatile fission products (VFPs) must a priori include a realistic description of fission-gas release and swelling. The steady-state and transient gas release and swelling subroutine, FASTGRASS, has been modified to include a mechanical description of behavior of VFPs (I, Cs, CsI, Cs2MoO4, and Cs2UO4). Phenomena modeled are the chemical reactions between the VFPs, VFP migration through the fuel, and VFP interaction with the noble gases. The paper will describe calculations performed with FASTGRASS to describe the releases of noble gases, I, Cs, and CsI from LWR fuel during steady-state and power-ramping conditions. Key issues that are addressed in the analysis are the effects of (a) VFP chemistry, (b) various assumptions concerning mechanisms of VFP migration through solid UO2, (c) fission-gas behavior, and (d) accident scenario on the chemical form of iodine and the rate of iodine release from water-reactor fuel. (author)

  9. Engineering and Economic Aspects of Mirror Machine Reactors with Direct Conversion

    International Nuclear Information System (INIS)

    Reactor design studies are presented based on the use of mirror confinement zones fed by neutral beam injectors and utilizing direct converters for charged-particle-energy recovery. Designs considered include Yin-Yang and axially symmetric coil configurations, D-T and D-3He fuel cycles, and net electrical outputs ranging from one hundred to one thousand megawatts, the latter being the base case. The operating power level of each reactor component is determined as a function of component efficiencies, Q (defined as the ratio of fusion power to trapped injected power), and other relevant variables. Then approximate cost-power scaling relationships are used to calculate component costs. Results for overall cost and system efficiency are presented as functions of Q for a variety of component efficiency sets. The results indicate that the D-T system with direct conversion is economically attractive for expected Q-values and component efficiencies. In comparison, the D-3He system is characterized by high sensitivity to system changes at expected Q-values, and very high component efficiencies are required in order to make it economical. The disadvantage of the D-T-system is that it is basically a heat engine and has little potential for overall system efficiencies greater than 45% at blanket temperatures usually considered. In contrast, D-3He has a potential for operating economically at system efficiencies greater than 80%. Such a system could be achieved if Q-values for D-3He near unity become possible and sufficient ingenuity in the design of efficient reactor components is exercised. (author)

  10. Fission product retention in the ACACIA (AdvanCed Atomic Cogenerator for Industrial Applications) reactor primary system

    International Nuclear Information System (INIS)

    The transport and deposition of fission products in the ACACIA (AdvanCed Atomic Cogenerator for Industrial Applications) high temperature reactor primary system are investigated. The study focuses on the behaviour calculated with the MELCOR computer code of 4 nuclides: Cs-137, Sr-90, Ag-110m, and I-131. After a ten-year operation period, the total activity of the released nuclides in the primary system is about 58 GBq. The highest activity is produced by Cs-137 (52 GBq), followed by I-131 (4 GBq), and Ag-110m (1.8 GBq). The contribution of Sr-90 is very low (1600 Bq). The highest activity is found in the precooler (56 GBq). The main reason is the condensation of the volatile species CsOH and CsI in this component. Other components with high activities are the recuperator (1.4 GBq) and the compressor (0.007 GBq). These components are contaminated by Ag-110m. The gas ducts in the energy conversion unit are contaminated by Ag-110m (0.043 GBq) and I-131 (0.011 GBq). Contamination as a consequence of a Loss Of Coolant Accident (LOCA) or a Loss of Flow Accident (LOFA) is difficult to estimate, because it depends on a number of phenomena. Under the assumption that 10 fuels elements are damaged, the activity is about 44 GBq. In this accident, Ag-110m produces most of the activity (31 GBq), followed by Cs-137 (12 GBq), and I-131 (1.4 GBq). Although the activity of Sr-90 is high (0.023 GBq), it is negligible compared to the activity of the other nuclides. Obviously, the activity is proportional to the number of damaged fuel elements. The distribution of the nuclides over the components is not considered in view of the large number of uncertainties. 9 refs

  11. MANTRA: An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    Science.gov (United States)

    Youinou, G.; Vondrasek, R.; Veselka, H.; Salvatores, M.; Paul, M.; Pardo, R.; Palmiotti, G.; Palchan, T.; Nusair, O.; Nimmagadda, J.; Nair, C.; Murray, P.; Maddock, T.; Kondrashev, S.; Kondev, F. G.; Jones, W.; Imel, G.; Glass, C.; Fonnesbeck, J.; Berg, J.; Bauder, W.

    2014-05-01

    This paper presents an update of an on-going collaborative INL-ANL-ISU integral reactor physics experiment whose objective is to infer the effective neutron capture cross sections for most of the actinides of importance for reactor physics and fuel cycle studies in both fast and epithermal spectra. Some fission products are also being considered. The principle of the experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation together with the neutron fluence will allow inference of effective neutron capture cross-sections in different neutron spectra.

  12. Fission density distribution measurement in vicinity of VVER-440 control assembly model in LR-0 reactor. Variant 4. Boron acid in moderator and control assembly coupler without Hf inserts

    International Nuclear Information System (INIS)

    The axial and radial fission product activity distributions in fuel assemblies surrounding a model of a partly inserted VVER-440 control rod (core fuelling variant 4) were measured on the LR-0 experimental reactor. The reactor core for the measurement consisted of 19 fuel assemblies of the VVER-440/LR-0 type

  13. Preliminary three-dimensional neutronics design and analysis of helium-cooled blanket for a multi-functional experimental fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    A multi-functional experimental fusion-fission hybrid reactor concept named FDS-MFX, which is based on viable fusion and fission technologies, has been proposed. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this paper, the design optimization for the layout and the size of high enriched uranium modules in later stage of uranium-fueled blanket has been performed. Finally, proposing a preliminary three-dimension neutronics design with maximum average Power Density (PDmax) 100 MW/ m3, loaded mass of the 235U 1000 kg and TBR (Tritium Breeding Ratio) 1.05. (authors)

  14. Civacuve analysis software for mis machine examination of pressurized water reactor vessels

    International Nuclear Information System (INIS)

    The product software CIVACUVE is used by INTERCONTROLE for the analysis of UT examinations, for detection, performed by the In-Service Inspection Machine (MIS) of the vessels of nuclear power plants. This software is based on an adaptation of an algorithm of SEGMENTATION (CEA CEREM), which is applied prior to any analysis. It is equipped with tools adapted to industrial use. It allows to: - perform image analysis thanks to advanced graphic tools (Zooms, True Bscan, 'contour' selection...), - backup of all data in a database (complete and transparent backup of all informations used and obtained during the different analysis operations), - connect PC to the Database (export of Reports and even of segmented points), - issue Examination Reports, Operating Condition Sheets, Sizing curves... - and last, perform a graphic and numerical comparison between different inspections of the same vessel. Used in Belgium and France on different kind of reactor vessels, CIVACUVE has allowed to show that the principle of SEGMENTATION can be adapted to detection exams. The use of CIVACUVE generates a important time gain as well as the betterment of quality in analysis. Wide data opening toward PC's allows a real flexibility with regard to client's requirements and preoccupations

  15. Charge distribution studies in the fast-neutron-induced fission of sup 2 sup 3 sup 2 Th, sup 2 sup 3 sup 8 U, sup 2 sup 4 sup 0 Pu and sup 2 sup 4 sup 4 Cm

    CERN Document Server

    Naik, H; Iyer, R H

    2003-01-01

    Charge distribution studies for heavy-mass fission products were carried out in the fast-neutron-induced fission of sup 2 sup 3 sup 2 Th, sup 2 sup 3 sup 8 U, sup 2 sup 4 sup 0 Pu and sup 2 sup 4 sup 4 Cm using radiochemical and gamma-ray spectrometric techniques. The width parameter(sigma sub Z /sigma sub A), the most probable charge/mass (Z sub P /A sub P), the charge polarization (DELTA Z) and the slope of charge polarization [ delta(DELTA Z)/delta A sup '] as a function of the fragment mass (A sup ') were deduced. The average charge dispersion parameter (left angle sigma sub Z right angle) and proton odd-even effect (delta sub p) were also obtained for these fissioning systems. The left angle sigma sub Z right angle and delta sub p values in the fissioning systems sup 2 sup 4 sup 1 Pu sup * and sup 2 sup 4 sup 5 Cm sup * were determined for the first time. The delta(DELTA Z)/delta A sup ' value is also determined for the first time in the fissioning systems sup 2 sup 3 sup 9 U sup * , sup 2 sup 4 sup 1 Pu...

  16. Fission product chemistry and aerosol behaviour in the primary circuit of a pressurised water reactor under severe accident conditions

    International Nuclear Information System (INIS)

    Three key accident sequences are considered covering a representative range of different environments of pressure, flow, temperature history and degree of zircaloy oxidation, and their principle thermal hydraulic and physical characteristics affecting chemistry behaviour are identified. Inventories, chemical forms and timing of fission product release are summarized together with the major sources of structural materials and their release characteristics. Chemistry of each main fission product species is reviewed from available experimental and/or theoretical data. Studies modelling primary circuit fission product behaviour are reviewed. Requirements for further study are assessed. (UK)

  17. Analysis of linear energy transfers and quality factors of charged particles produced by spontaneous fission neutrons from 252Cf and 244Pu in the human body

    International Nuclear Information System (INIS)

    Absorbed doses, linear energy transfers (LETs) and quality factors of secondary charged particles in organs and tissues, generated via the interactions of the spontaneous fission neutrons from. 252Cf and. 244Pu within the human body, were studied using the Particle and Heavy Ion Transport Code System (PHITS) coupled with the ICRP Reference Phantom. Both the absorbed doses and the quality factors in target organs generally decrease with increasing distance from the source organ. The analysis of LET distributions of secondary charged particles led to the identification of the relationship between LET spectra and target-source organ locations. A comparison between human body-averaged mean quality factors and fluence-averaged radiation weighting factors showed that the current numerical conventions for the radiation weighting factors of neutrons, updated in ICRP103, and the quality factors for internal exposure are valid. (authors)

  18. Investigation of mass and nuclear charge distributions in a fission induced by 3 MeV neutrons for some fissile nuclei

    International Nuclear Information System (INIS)

    After a presentation of the phenomenon of fission (liquid droplet model, microscopic model, Strutinski model, static approach to the scission point with the Fong statistical model and with the Wilkins thermodynamic model), this research thesis presents an experimental installation with its irradiation systems, its measurement assembly, its measurement process (rare gas emission). The author then describes the methods used to determine efficiencies: charge distributions within an isobaric chain, efficiency determination principle, choice of experimental parameters, test with Uranium 235. Experimental results are then presented and discussed in terms of mass distribution and of charge distribution for various uranium isotopes (235, 238 and 232). They are discussed with respect to the Wilkins model, to the pair breakage model, and to the calculation of the average number of neutrons emitted by different fissile systems

  19. Thermodynamics mechanisms of fission product retention in nuclear plants illustrated by the properties of the HTR reactor

    International Nuclear Information System (INIS)

    Starting from the first law of thermodynamics, the theoretical principles for the description of interactions between fission products and other materials are derived step by step, using fundamental terms such as phase equilibria, mixtures and solutions. Thereafter, the concepts of Onsager's theory of irreversible thermodynamics are introduced. They serve as an example of modelling fission product transport with special respect to thermochemical properties. In the last chapter real technical concepts for fission product retention are evaluated using thermodynamic criteria. A fine distinction is performed between barrier-, filter- and sinkmechanisms for retention-purposes. One important result is, that a barrier-concept alone doesn't meet the challenge of nuclear power operation without the probability of hazardous accidents. The work is finished by a proposal to improve the fission product retention capabilities of HTR fuel-elements in combination with a coating of the fuel-pebbles. (orig./DG)

  20. Contamination of the air and other environmental samples of the Ulm region by radioactive fission products after the accident of the Chernobyl reactor

    International Nuclear Information System (INIS)

    Since April 30, 1986, the radioactivity of the fission products released by the accident of the Chernobyl reactor has been measured in the air of the city of Ulm. The airborne dust samples were collected with flow calibrated samplers on cellulose acetate membrane filters and counted with a high resolution gamma ray spectrometer. Later on, the radioactivity measurements were expanded to other relevant environmental samples contaminated by radioactive atmospheric precipitates including grass, spruce needles, mosses, lichens, various kinds of food, drinking water, asphalt and concrete surface layers, municipal sewage sludge and sewage sludge ash. This paper reports the obtained results. (orig.)

  1. Fast-Mixed Spectrum Reactor progress report. Results of the FMSR Benchmark calculations and an assessment of current fission product libraries

    International Nuclear Information System (INIS)

    As part of the Initial Feasibility Study of the Fast Mixed Spectrum Reactor, a series of benchmark calculations were made to determine the sensitivity of the physics analysis to differences in methods and data. Argonne National Laboratory (ANL), the Massachusetts Institute of Technology (MIT), and Oak Ridge National Laboratory (ORNL) were invited to participate with Brookhaven National Laboratory in the analysis of a FMSR model prescribed by BNL. Detailed comparisons are made including a comprehensive study on the adequacy of the fission product treatments

  2. Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on europium: small scale production of no-carrier-added 153Sm in a nuclear reactor

    International Nuclear Information System (INIS)

    Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, 2n), (n, p) and (n, α) reactions on isotopes of europium. Our measurements constitute the first systematic studies. Of special interest was the investigation of 153Eu(n, p)153Sm reaction for the production of no-carrier-added 153Sm in a nuclear reactor. Using 100% enriched 153Eu target, 97.21 MBq 153Sm per batch can be produced which is, however, not sufficient for medical application. (orig.)

  3. 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)

  4. Determination of isobar composition and yields of 239Pu fission-products by thermal neutrons

    International Nuclear Information System (INIS)

    On the research nuclear reactor WWR-SM of INP Uz AS by means of mass-spectrometer the heavy fission-products of 239Pu nuclei induced by thermal neutrons are measured in ranges of mass Ai = 125 -157, kinetic energies Ek = 45 - 87 MeV and effective ionic charges z* = 18 - 30. 102 isobar nuclei in composition of the measured fission-products, also the partial yields of the each element giving the contribution to formation of a total yield of heavy fission-product with mass Ai are defined. (authors)

  5. Extraction of neutron-rich fission products from a nuclear reactor for ground state studies: commissioning of the online-coupling at TRIGA-SPEC

    International Nuclear Information System (INIS)

    The mass spectrometer TRIGA-TRAP and the laser spectroscopy TRIGA-LASER setup, forming the TRIGA-SPEC experiment, are installed at the research reactor TRIGA Mainz in order to perform high-precision measurements of the ground state properties of short-lived neutron-rich radionuclides. Such measurements allow testing the predictive power of nuclear mass models and support astrophysical nucleosynthesis calculations. The extraction of these nuclei for both experiment branches is achieved by using an aerosol-based gas-jet system to transport fission products from an actinide target located inside the reactor to an external high-temperature surface ion source. TRIGA-SPEC will shortly go online, already having recorded a cyclotron resonance of an ion produced in the source. The commissioning of the online-coupling involving a separator magnet, a radiofrequency quadrupole cooler/buncher, and a pulsed drift tube will be presented.

  6. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle; 1: l'atome. 2: la radioactivite. 3: l'homme et les rayonnements. 4: l'energie. 5: l'energie nucleaire: fusion et fission. 6: le fonctionnement d'un reacteur nucleaire. 7: le cycle du combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  7. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  8. Measurement of average cross section for Pa-233 (n, 2n) Pa-232 reaction to neutrons with fission-type reactor spectrum

    International Nuclear Information System (INIS)

    Among some nuclides concerning thorium fuel cycle, the reaction cross sections of Pa-233 should be thoroughly investigated because of its relatively long life of 27 days half life. In the present works, the average cross section for Pa-233(n,2n)Pa-232 reaction, which has been considered to contribute to the production of troublesome concomitant U-232, was initially measured using the Pa-233 specimen as pure as possible followed by the re-irradiation in a fission-type neutron spectrum. The purest Pa-233 was produced from the first thermal neutron irradiation of Th-oxide, which was selected from the viewpoint of low Th-230 content to avoid the production of bothering Pa-231 having a large cross section for thermal neutron capture reaction. The chemically isolated Pa-233 was immediately re-irradiated with reactor neutrons having fission-type reactor spectrum in KUR, along with some flux monitors for fast neutrons. After completely decaying out Pa-233 to U-233, the chemical purification of uranium was performed and the resultant uranium isotopes were analysed with an alpha-spectrometry. By using the activity ratios of U-232/U-233, the objective cross section was evaluated to be 52.1 mbarn with an estimated overall experimental error of 10 % after correcting the inevitable bypath reaction by small amount of Pa-231 content. (author)

  9. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    International Nuclear Information System (INIS)

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  10. Health effects of low dose exposure to fission products from Chernobyl and the Fermi nuclear reactor in the population of the Detroit metropolitan area

    International Nuclear Information System (INIS)

    The present paper describes the results of the exposure of a very large population in the Detroit, Michigan, area to fallout from Chernobyl measured in 1986, followed by the reported releases from the start-up of the Fermi-II nuclear plant in 1988 located 20 miles from the city that receives its drinking water from Lake St. Clair downwind to the north-east of the plant. Due to the prior existence of a local cancer registry for a total population of about 4 million, and the availability of reliable public-heath statistics by age, race and sex, combined with the absence of an accident known to produce population movement and stress, highly significant rises and declines of the incidence of early childhood leukemia and other cancers could be related both geographically and temporally to the observed rises and declines of fission products in the milk as well as releases from the reactor. Furthermore, surprisingly rapid rises in the incidence of breast cancer also took place in Monroe County where the reactor is located and in Macomb County downwind on Lake St. Clair to the northeast, presumably due to weakening of the immune defenses by the mix of fission products not seen so rapidly after exposure in the case of external X-rays or gamma rays. For Michigan as a whole, for which incidence of thyroid cancer at all ages combined became available after 1985, rapid rises were observed after Chernobyl and the start of the Fermi plant, using as rapidly as in the case of Belarus and Connecticut. Additionally, highly significant synchronous rises in low birth weight, infant mortality, fetal deaths, asthma and infectious disease mortality were also observed consistent with the known action of bone-seeking fission products on the immune system, following reported nuclear tests, nuclear accidents and the start-up of the Fermi plant. (orig.)

  11. Fission Fragments Discriminator

    International Nuclear Information System (INIS)

    Nuclear fission reaction between Uranium-235 nucleus and thermal neutron caused the high energy fission fragments with uncertainly direction. The particle direction discrimination was determined. The 2.5 x 3.0 mm2 polyethylene gratings with 1-6 mm thickness were used. The grating was placed between uranium screen that fabricated from ammonium-diurinate compound and polycarbonate nuclear track film recorder irradiated by neutron from Thai Research Reactor (TRR-1/M1) facility. The nuclear track density was inversely with grating thickness. It's only fission fragments normal to uranium screen pass through film recorder when grating thickness was 4-6 mm

  12. A feasibility study of coolant void detection in a lead-cooled fast reactor using fission chambers

    OpenAIRE

    Wolniewicz, Peter

    2012-01-01

    One of the future reactor technologies defined by the Generation-IV International Forum (GIF) is the Lead-Cooled Fast Reactor (LFR). An advantage with this reactor technology is that steam production is accomplished by means of heat exchangers located within the primary reactor vessel, which decreases costs and increases operational safety. However, a crack in a heat exchanger tube may create steam (void) into the coolant and this process has the potential to introduce reactivity changes, whi...

  13. Automatically controlled electric motor with inductive rheostat for the drive system of a coal charging machine

    Energy Technology Data Exchange (ETDEWEB)

    Orekhovskii, V.P.; Tarasov, S.P.; Ivashchenko, V.A.; Ivanov, S.I.; Starichenko, N.V. (KB Koksokhimmash Giprokoksa (USSR))

    1989-07-01

    Discusses design and operation of start-up systems for drives of systems for coal charging to coke ovens. Two systems are comparatively evaluated: asynchronous motors with start-up resistors or with inductive rheostats. Use of inductive rheostats simplifies design of drive systems, increases operational reliability (as the number of elements characterized by lowest service life and highest failure rate declines). Replacing start-up resistors with rheostats leads to a decrease in speed of coal charging. Starting acceleration decreases, acceleration distance increases, braking distance increases and braking time also increases. Advantages associated with increased reliability of the charging system compensate reduced loading speed.

  14. Detection of the weak neutral current using fission anti ν/sub e/ on deuterons with concurrent measurement of the charged current branch

    International Nuclear Information System (INIS)

    The target consists of 268 kg of extremely pure (99.85%) heavy water (D2O), contained in a cylindrical stainless steel tank 122 cm in height and 54 cm in diameter. This target is surrounded by a lead and cadmium shield and immersed in a 2200 liter liquid scintillator anticoincidence detector. This system is a well-shielded environment. The center of the detector is located 11 meters from the center of the reactor core in an electron antineutrino flux of 2.5 x 1013 anti ν/sub e//cm2-s. Immersed in the target are ten 5.08-cm-diameter 3He-filled gas proportional counters, which detect the neutron via the reaction 3He + n → p + 3H + 773 keV. The system has been determined to have an overall neutron detection efficiency of 0.32 +- 0.02. The data are taken with a combination of scalers, a pulse height analyzer, and oscilloscope traces. Single, double, and triple neutron capture events are recorded with the reactor on and off. Data have been accumulated for 104 live days reactor on, and 72 live days reactor off for the charged-current reaction and 52 live days reactor on and 34 live days reactor off for the neutral-current reaction. The measured neutral-current cross section is (5.0 +- 0.8) x 10-45cm2/anti ν/sub e/, consistent with the Weinberg-Salam model. The charged-current reaction cross section is (1.5 +- 0.4) x 10-45 cm2/ν/sub e/, in fair agreement with expectation. From the N.C. cross section a value of the square of the isovector axial-vector coupling constant is deduced to be β2 = 1.0 +- 0.15

  15. Performance of the fissionTPC and the Potential to Advance the Thorium Fuel Cycle

    Science.gov (United States)

    Towell, Rusty; Niffte Collaboration

    2014-09-01

    The NIFFTE fission Time Projection Chamber (fissionTPC) is a powerful tool that is being developed to take precision measurements of neutron-induced fission cross sections of transuranic elements. During the last run at the Los Alamos Neutron Science Center (LANSCE) the fully instrumented TPC took data for the first time. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow cross section systematics to be controlled at the level of 1%. The fissionTPC performance from this run will be shared. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors including the abundance of thorium verses uranium, minimizing radioactive waste, improved reactor safety, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium fueled nuclear reactors will also be discussed.

  16. Improved Calculation of Thermal Fission Energy

    OpenAIRE

    Ma, X. B.; Zhong, W. L.; Wang, L. Z.; Y. X. Chen; Cao, J

    2012-01-01

    Thermal fission energy is one of the basic parameters needed in the calculation of antineutrino flux for reactor neutrino experiments. It is useful to improve the precision of the thermal fission energy calculation for current and future reactor neutrino experiments, which are aimed at more precise determination of neutrino oscillation parameters. In this article, we give new values for thermal fission energies of some common thermal reactor fuel isotopes, with improvements on three aspects. ...

  17. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Postirradiation heating tests of TRISO-coated UO2 particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of 85Kr, 110mAg, 134Cs, 137Cs, and 154Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of 110mAg, 134Cs, 137Cs, and 154Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations

  18. Radiolytic production of chemical fuels in fusion reactor systems

    International Nuclear Information System (INIS)

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered

  19. Fission product detection

    International Nuclear Information System (INIS)

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells

  20. Resistance-to-wear testing of metallic machine components by charged particle surface activation

    International Nuclear Information System (INIS)

    Surface activation, commonly known as Thin Layer Activation (TLA), is an ion beam technique generally employed for wear and/or corrosion monitoring in industrial plants, especially related to applications in engine and machine construction industries. For an efficient industrial application of this technique, one must foresee some specific irradiation set-ups as well as the measuring method, having in mind the diversity of machine components which might be subjects of such studies and the tribological phenomena involved. Some of these aspects are mentioned in the paper. In principle, the TLA-based measuring methods exploit the correlation between the loss of material owing to wear and the resulting changes in radioactivity. Two such measuring methods are derived; they will be reviewed briefly in the paper and illustrated by wear diagrams obtained for several parts of running machines such as piston-rings, linear cylinders and bearing crankshafts, together with a study of characteristics of different (unspecified) lubricant oils. An example where a TLA-based method has been successfully applied, for material couple characterisation on Timken testing stands, is also presented. The paper ends with a typical remnant measuring method application executed for the Romanian Railway Company. (Author)

  1. Fuels for space nuclear power systems. 3. Innovative Semi-spherical Pb-Hf-Cu Shield for a Fissioning Plasma Core Reactor

    International Nuclear Information System (INIS)

    This study investigated the shielding materials and requirements for a fissioning plasma core reactor (FPCR) with a magnetohydrodynamic (MHD) power conversion system for multimegawatt space power and propulsion applications. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. This concept is under study for its potential to provide space power at a low specific mass of 3 with a length-to-diameter ratio of one was selected. This design based on earlier gas core reactor studies incorporates a 50-cm BeO radial reflector with additional 25-cm-thick BeO disk-shaped reflectors at the top and bottom of the cylindrical core. Liquid hydrogen tanks for propulsion and refrigeration were modeled between the reactor/power generation complex and the payload/habitable regions of the vessel or space station and lying along the boom, which can be from 30 to 60 m in length. Although the liquid hydrogen is not very dense (∼0.1 g/cm3), there is a considerable amount present (50 t is commonly referenced). A model of this system was developed in the MCNP-4C general Monte Carlo code, which was used to calculate the dose rate at various distances from the power-generating system. Sources of both fission neutrons (prompt and delayed) and gamma rays (prompt and decay) from fission were modeled. The neutron sampling distribution was taken as a Watt fission spectrum. The energy distribution of gamma rays from fission was taken from Ref. 1 and consists of a total average of 12.15 gamma rays per fission with ∼70% < 1 MeV and 27% between 1 and 3 MeV. Various shield designs were modeled, and corresponding dose rates were calculated. A criterion of <10 rems/h to the payload module was established for all shield designs. It was assumed that for a manned station or vessel, additional shielding would be

  2. Delayed fission

    International Nuclear Information System (INIS)

    Delayed fission is a nuclear process that couples beta decay and fission. In the delayed fission process, a parent nucleus undergoes beta decay or electron capture and thus populates excited states in the daughter nucleus. This review covers experimental methods for detecting and measuring delayed fission. Experimental results (ECDF activities and beta-DF activities) and theory are presented. The future prospects for study of delayed fission are discussed. 50 refs., 8 figs., 2 tabs

  3. Delayed Neutrons and Photoneutrons from Fission Products

    International Nuclear Information System (INIS)

    Delayed neutrons: Most studies of the delayed neutrons from fission have involved analysis of the kinetic behaviour of fusion chain- reacting systems, analysis of the gross neutron decay (resolved into six groups with approximate half-lives of 0.2, 0.5, 2, 6, 22 and 55 s) and some measurements of the neutron spectra (the energies extendfrom 0.1 to 1.2 MeV, peaking in the range 0.2 to 0.5 MeV). Rapid separations of fission-produced halogens have indicated seven isotopes (Br87,88,89,90 and I137,138,139). and rare gas analysis has indicated 1.5-s Kr and 6-s Rb as definite delayed neutron precursors. These identified precursors account for some 80% of the total delayed neutron yields. Theoretical predictions of possible precursors point to a few tens of such nuclides to be found mainly in regions just above closed neutron shells. Total neutron yields are observed to increase with mass number and decrease with atomic number of the fissioning nuclide. Yields are nearly independent of the energy of the incident fissioning neutron at energies up to several MeV. In this range observed group yields,-especially of the long-lived precursors, ate in fairly good agreement with fission mass and charge distributions, and calculated neutron emission probabilities. . Further detailed studies of delayed neutron precursors (particularly in the difficult short half-life region) require development of ultra-fast radiochemical separation procedures (or on-line isotope separation) and fast neutron spectroscopy of high resolution and efficiency. Photoneutrons; A knowledge of the intensities and gamma-ray spectra of fission products is of practical importance in reactor technology particularly with respect to gamma heating, shielding and radiation effects. Gamma-rays of energies greater than 2.23 and 1.67 MeV cause emission of photoneutrons from deuterium and beryllium respectively, and are important in the kinetics of heavy water and beryllium-moderated reactors. The rate of photoneutron

  4. Measurement of mass and isotopic fission yields for heavy fission products with the LOHENGRIN mass spectrometer

    International Nuclear Information System (INIS)

    In spite of the huge amount of fission yield data available in different libraries, more accurate values are still needed for nuclear energy applications and to improve our understanding of the fission process. Thus measurements of fission yields were performed at the mass spectrometer Lohengrin at the Institut Laue-Langevin in Grenoble, France. The mass separator Lohengrin is situated at the research reactor of the institute and permits the placement of an actinide layer in a high thermal neutron flux. It separates fragments according to their atomic mass, kinetic energy and ionic charge state by the action of magnetic and electric fields. Coupled to a high resolution ionization chamber the experiment was used to investigate the mass and isotopic yields of the light mass region. Almost all fission yields of isotopes from Th to Cf have been measured at Lohengrin with this method. To complete and improve the nuclear data libraries, these measurements have been extended in this work to the heavy mass region for the reactions 235U(nth,f), 239Pu(nth,f) and 241Pu(nth,f). For these higher masses an isotopic separation is no longer possible. So, a new method was undertaken with the reaction 239Pu(nth,f) to determine the isotopic yields by spectrometry. These experiments have allowed to reduce considerably the uncertainties. Moreover the ionic charge state and kinetic energy distributions were specifically studied and have shown, among others, nanosecond isomers for some masses. (author)

  5. An economics method for symbiotic fusion-fission electricity generation systems

    International Nuclear Information System (INIS)

    A self-consistent analytical methodology for evaluating the economic incentives for symbiotic electricity generation systems that consist of fusion breeder reactors and supported fission converter reactors is developed. This methodology employs a discounted cash flow analysis of breeder and converter direct operating costs and indirect capital costs, as well as a novel treatment of fissile inventory charges. Three figures of merit are emphasized: the levelized cost of electricity generated by the symbiotic system; the levelized cost of fuel exchanged by the breeder and converter reactors in the system; and the equivalent cost of fuel to produce the same levelized electricity cost in a alternatively fueled converter reactor. A fission converter operating on the current once-through fuel cycle is a special case of the above. The method is equally applicable to symbiotic systems that utilize spallation accelerator breeder reactors

  6. Measurements of the yields of the light fission products from the reaction 233U(nsub(th),f) by a ionization chamber

    International Nuclear Information System (INIS)

    The aim of this thesis was to develop a new measuring apparature and measuring method which allows to study together with the mass separator 'Lohengrin' at the high flux reactor in Grenoble in realizable measurement times detailedly the unknown mass, nuclear charge, and energy distributions of the fission products resulting from the fission of 233U with thermal neutrons. First the yields and the energy distributions of the masses, thereafter the yields and the energy distributions of the isobaric nuclear charges of the light fission products in the mass range 79<=Asub(L)<=106 are measured. The measuring method for the determination of the mass yields consists of a energy measurement of the fission products separated in the mass separator by a ionization chamber. The isobaric nuclear charges and their yields are determined by the nuclear-charge-specific energy-loss method from the residual-energy spectra behind an absorber. (orig./HSI)

  7. Fission gas release (FGASRL)

    International Nuclear Information System (INIS)

    During irradiation of water reactor fuel rods, gaseous fission products are produced in the fuel and are slowly released to various voipd volumes in the fuel rods. The released fission gases degrade the initial fill gas thermal conductivity and thus change the thermal response of the fuel rods. Moreover, fuel rod internal pressure is increased so that the cladding mechanical response is affected. The fission gas release subcode FGASRL is intended for use in analytical codes which predict water reactor fuel pin behavior. The development effort was directed primarily at improving code predictions of the gas release model used in FRAP-S3 which overpredicts release of fuels irradiated at relatively low operating temperatures and therefore small gas release fractions. The fission gas release subcode (FGASRL) presented in the report describes a two-step gas release process: (a) fission gas release from fuel grains to the grain boundaries, and (b) fission gas release from the grain boundaries to internal free volume of the fuel pin

  8. A Covariance Generation Methodology for Fission Product Yields

    Directory of Open Access Journals (Sweden)

    Terranova N.

    2016-01-01

    Full Text Available Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1 no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation, developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  9. A Covariance Generation Methodology for Fission Product Yields

    Science.gov (United States)

    Terranova, N.; Serot, O.; Archier, P.; Vallet, V.; De Saint Jean, C.; Sumini, M.

    2016-03-01

    Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1) no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM) implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation), developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  10. Void Reactivity Effects in the Second Charge of the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    The reactivity effect of voids caused by boiling inside the coolant channels in the second fuel charge of the Halden Boiling Heavy Water Reactor has been measured both in void-simulated zero-power experiments and under actual power conditions. The void-simulated experiments consisted of measuring the reactivity effect of introducing void columns inside thin-walled tubes to various depths. The tubes were placed at different positions between die stringers in a single 7-rod cluster element practically identical with the normal second-charge fuel elements. This experiment enables an investigation of the reactivity dependence upon void fraction, and also the reactivity dependence of steam-bubble position in the coolant channel. The experiment was carried out in the Norwegian zero-power facility NORA, with a core consisting of 36 second-charge elements and with a lattice geometry identical to the one in HBWR. The temperature dependence of the void effect was investigated in a zero-power experiment with the 100 fuel-element core of HBWR. In a single fuel element the water level inside the coolant channel was depressed to various depths, and the reactivity effect of this perturbation was measured at different temperatures in the temperature interval 50°C-220°C. The power void reactivity has been measured in HBWR as a function of nuclear power at different moderator temperatures between 150°C and 230°C at powers up to about 16 MW at the highest temperature. The power-void reactivity coefficient is an important quantity in determining the dynamic behaviour of a boiling- water reactor. The theoretical determination of this quantity is, however, complicated by the fact that knowledge about the void distribution in the core is required. The detailed power-void distribution is not easily amenable to experimental determination, and accordingly the void-simulated experiments represent a better case for testing the reactor physics calculation of void effects. Preliminary

  11. Activation analysis and waste management for blanket materials of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket

  12. Chromosomal mutation by fission neutrons and X-rays in higher plants. A review on results of the joint research program utilizing Kinki University reactor

    International Nuclear Information System (INIS)

    We have studied the efficiency of fission neutrons from the nuclear reactor of Kinki University (UTR-KINKI) and X-rays to chromosomes of higher plants for over 20 years. In this review, we described the development of bio-dosimeter using hyper-sensibility of germinating onion roots for irradiation, the analysis of chromosome structure in Haplopappus gracilis (Asteraceae), with the special reference of latent centromeres and survived telomeres throughout chromosomal evolution, the experimental studies on the induction of chromosomal rearrangement in Zebrina pendula (Commelinaceae), the behavior of chromosome fragments with non-localized centromeres in Carex and Eleocharis (Cyperaceae), and the possibility as a bio-dosimeter of pollen mother cells of Tradescantia paludosa (Commelinaceae) for the detection of low-dose radiation. (author)

  13. MAFF - the Munich accelerator for fission fragments

    International Nuclear Information System (INIS)

    At the new high flux reactor FRM-II in Munich the accelerator MAFF (Munich Accelerator for Fission Fragments) is under design. In the high neutron flux of 1014 n/cm2 s up to 1014 neutron-rich fission fragments per second are produced in the 1 g 235 U target. Ions with an energy of 30 keV are extracted from the ion source. In the mass separator two isotopes can be selected. One of the beams is used for low energy experiments, the other one is injected into an ECRIS (or EBIS) for charge breeding to a q/A ≥ 0.16. A gas filled RFQ cooler is used for emittance improvement. The subsequent LINAC delivers beams with an energy ranging from 3.7 MeV/u to 5.9 MeV/u. New IH structures are being developed at the Munich tandem laboratory. A small storage ring is planned in a further stage to recycle the fission fragments. A thin target foil can be placed into this ring, e.g., for synthesis of super-heavy elements. The through-going beam tube has been installed in the heavy water tank of the reactor. Tests of the target ion source in a special oven to test long term stability and safety tests were in progress. (author)

  14. The utilization of 10Mw research reactor in Tashkent

    International Nuclear Information System (INIS)

    We present the short review of basic and applied research as well as data on the mass production of reactor isotopes with use of 10Mw water-water research reactor of the Institute of Nuclear Physics of Uzbekistan Academy of Sciences in Tashkent. Despite the relatively long time of operation (since 1959) the several projects on the modernization of machine have been done. The reactor operates more than 5000 hours a year and serves also for elemental analysis (neutron activation), radiochemistry, radiation hardness and fission products studies as well as for changing the properties of optical and semiconductor materials. Until 1997 reactor was operating with use of highly enriched fuel (90% of enrichment) and starting from the middle of 1997 it has been converted to use 36% enrichment fuel. In the second half of 2007 the preparatory works on the full conversion to the 19.7% enrichment fuel should be completed. We also present the results of our experience in sending highly enriched spent fuel back to country-origin (Russia) for first time in last 16 years. The external one-arm spectrometer of secondary fission products made it possible a study of properties (the charge, angular and momentum distributions) of fission products which revealed quite interesting features inconsistent with some standard models of fusion. The special method has been developed to carry out elemental analysis which once applied, for example, to pure metals allows to determine impurities with concentration up to 10-10 % (for almost 60 elements simultaneously). The reactor is also using for changing the properties of some materials like semiconductors, ceramics and natural crystals bringing their quality to market demands. The main activity of our reactor is the mass production of isotopes for medical and scientific needs. In particular, the isotopes like Tc-99m (from irradiation of Mo wires), P-32, P-33, I-125, I-131, S-35, Au-186, Sr-89, Fe-55 and some others are producing at the level of

  15. Macrocluster desorption effect caused by single MCI: charges of gold clusters (2-20 nm) desorbed due to electronic processes induced by fission fragment bombardment in nanodispersed gold targets

    International Nuclear Information System (INIS)

    In this work the charge state of the negatively charged gold nanocluster ions (2-20 nm) that were desorbed from nanodispersed gold islet targets by 252Cf fission fragments via electronic processes is studied. Mean cluster charge was calculated as a ratio of mean cluster mass to mean mass-to-charge ratio . Cluster masses were measured by means of a collector technique employing transmission electron microscopy and scanning force microscopy, while m/q was measured by means of a tandem TOF-spectrometer. It is shown that the nanocluster ions are mostly multiply charged (2-16e) and the charge increases non-linearly with the cluster size. The results are discussed

  16. Nuclear fission

    International Nuclear Information System (INIS)

    The nuclear fission process is pedagogically reviewed from a macroscopic-microscopic point of view. The Droplet model is considered. The fission dynamics is discussed utilizing path integrals and semiclassical methods. (L.C.)

  17. Compact Reactor

    International Nuclear Information System (INIS)

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

  18. Hot cell works and related irradiation tests in fission reactor for development of new materials for nuclear application

    International Nuclear Information System (INIS)

    Present status of research works in Oarai Branch, Institute for Materials Research, Tohoku University, utilizing Japan Materials Testing Reactor and related hot cells will be described.Topics are mainly related with nuclear materials studies, excluding fissile materials, which is mainly aiming for development of materials for advanced nuclear systems such as a nuclear fusion reactor. Conflict between traditional and routined procedures and new demands will be described and future perspective is discussed. (author)

  19. Perspective on the fusion-fission energy concept

    International Nuclear Information System (INIS)

    A concept which has potential for near-term application in the electric power sector of our energy economy is combining fusion and fission technology. The fusion-fission system, called a hybrid, is distinguished from its pure fusion counterpart by incorporation of fertile materials (uranium or thorium) in the blanket region of a fusion machine. The neutrons produced by the fusion process can be used to generate energy through fission events in the blanket or produce fuel for fission reactors through capture events in the fertile material. The performance requirements of the fusion component of hybrids is perceived as being less stringent than those for pure fusion electric power plants. The performance requirements for the fission component of hybrids is perceived as having been demonstrated or could be demonstrated with a modest investment of research and development funds. This paper presents our insights and observations of this concept in the context of why and where it might fit into the picture of meeting our future energy needs. A bibliography of hybrid research is given

  20. Depth-charge static and time-dependence perturbation/sensitivity system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code block for both static and time-dependence perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Labortary. The DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analysis of realistic multidimensional reactor models