Analysis of an homogeneous solution reactor for 99 Mo production
International Nuclear Information System (INIS)
Weir, A.; Lopasso, E.; Gho, C.
2007-01-01
The 99m Tc is the more used radioisotope in nuclear medicine, used in 80% of procedures of nuclear medicine in the world. This is due to their characteristics practically ideal for the diagnostic. The 99m Tc is obtained by decay of the 99 Mo, which can produce it by irradiating enriched targets in 98 Mo, or as fission product, irradiating uranium targets or by means of homogeneous solution reactors. The pattern of the used reactor in the neutron analysis possesses a liquid fuel composed of uranyl nitrate dissolved in water with the attach of nitric acid. This solution is contained in a cylindrical recipient of stainless steel reflected with light water. The reactor is refrigerated by means of an helicoidal heat exchanger immersed in the fuel solution. The heat of the fuel is removed by natural convection while the circulation of the water inside the exchanger is forced. The control system of the reactor consists on 6 independent cadmium bars, with followers of water. An auxiliary control system can be the level of the fuel solution inside container tank, but it was not included in the pattern in study. One studies the variations of the reactivity of the system due to different phenomena. An important factor during the normal operation of the reactor is the variation of temperature taking to a volumetric expansion of the fuel and ghastly effects in the same one. Another causing phenomenon of changes in the reactivity is the variation of the concentration of uranium in the combustible solution. An important phenomenon in this type of reactors is the hole fraction in the nucleus I liquidate due to the radiolysis and the possible boil of the water of the combustible solution. Some of the possible cases of abnormal operation were studied as the lost one of coolant in the secondary circuit of the heat exchanger, the introduction and evaporation of water in the nucleus. The reactivity variations were studied using the codes of I calculate MCNP, WIMS and TORT. All the
HOMOGENEOUS NUCLEAR POWER REACTOR
King, L.D.P.
1959-09-01
A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.
International Nuclear Information System (INIS)
2008-09-01
Technetium-99m ( 99m Tc), the daughter of Molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. It accounts for over twenty-five million medical procedures each year worldwide, comprising about 80% of all radiopharmaceutical procedures. 99 Mo is mostly prepared by the fission of uranium-235 targets in a nuclear reactor with a fission yield of about 6.1%. Currently over 95% of the fission product 99 Mo is obtained using highly enriched uranium (HEU) targets. Smaller scale producers use low enriched uranium (LEU) targets. Small quantities of 99 Mo are also produced by neutron activation through the use of the (n, γ) reaction. The concept of a compact homogeneous aqueous reactor fuelled by a uranium salt solution with off-line separation of radioisotopes of interest ( 99 Mo, 131 I) from aliquots of irradiated fuel solution has been cited in a few presentations in the series of International Conference on Isotopes (ICI) held in Vancouver (2000), Cape Town (2003) and Brussels (2005) and recently some corporate interest has also been noticeable. Calculations and some experimental research have shown that the use of aqueous homogeneous reactors (AHRs) could be an efficient technology for fission radioisotope production, having some prospective advantages compared with traditional technology based on the use of solid uranium targets irradiated in research reactors. This review of AHR status and prospects by a team of experts engaged in the field of homogeneous reactors and radioisotope producers yields an objective evaluation of the technological challenges and other relevant implications. The meeting to develop this report facilitated the exchange of information on the 'state of the art' of the technology related to homogeneous aqueous solution nuclear reactors, especially in connection with the production of radioisotopes. This publication presents a summary of discussions of a consultants meeting which is followed by the technical
International Nuclear Information System (INIS)
Chakraverty, S.; Nayak, S.
2013-01-01
Highlights: • Uncertain neutron diffusion equation of bare square homogeneous reactor is studied. • Proposed interval arithmetic is extended for fuzzy numbers. • The developed fuzzy arithmetic is used to handle uncertain parameters. • Governing differential equation is modelled by modified fuzzy finite element method. • Fuzzy critical eigenvalues and effective multiplication factors are investigated. - Abstract: The scattering of neutron collision inside a reactor depends upon geometry of the reactor, diffusion coefficient and absorption coefficient etc. In general these parameters are not crisp and hence we get uncertain neutron diffusion equation. In this paper we have investigated the above equation for a bare square homogeneous reactor. Here the uncertain governing differential equation is modelled by a modified fuzzy finite element method. Using modified fuzzy finite element method, obtained eigenvalues and effective multiplication factors are studied. Corresponding results are compared with the classical finite element method in special cases and various uncertain results have been discussed
Homogenization theory in reactor lattices
International Nuclear Information System (INIS)
Benoist, P.
1986-02-01
The purpose of the theory of homogenization of reactor lattices is to determine, by the mean of transport theory, the constants of a homogeneous medium equivalent to a given lattice, which allows to treat the reactor as a whole by diffusion theory. In this note, the problem is presented by laying emphasis on simplicity, as far as possible [fr
Assembly homogenization techniques for light water reactor analysis
International Nuclear Information System (INIS)
Smith, K.S.
1986-01-01
Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed and numerical examples given. The mathematical foundations for homogenization methods are outlined. Two methods, Equivalence Theory and Generalized Equivalence Theory which are theoretically capable of eliminating homogenization error are reviewed. Practical means of obtaining approximate homogenized parameters are presented and numerical examples are used to contrast the two methods. Applications of these techniques to PWR baffle/reflector homogenization and BWR bundle homogenization are discussed. Nodal solutions to realistic reactor problems are compared to fine-mesh PDQ calculations, and the accuracy of the advanced homogenization methods is established. Remaining problem areas are investigated, and directions for future research are suggested. (author)
Cross section homogenization analysis for a simplified Candu reactor
International Nuclear Information System (INIS)
Pounders, Justin; Rahnema, Farzad; Mosher, Scott; Serghiuta, Dumitru; Turinsky, Paul; Sarsour, Hisham
2008-01-01
The effect of using zero current (infinite medium) boundary conditions to generate bundle homogenized cross sections for a stylized half-core Candu reactor problem is examined. Homogenized cross section from infinite medium lattice calculations are compared with cross sections homogenized using the exact flux from the reference core environment. The impact of these cross section differences is quantified by generating nodal diffusion theory solutions with both sets of cross sections. It is shown that the infinite medium spatial approximation is not negligible, and that ignoring the impact of the heterogeneous core environment on cross section homogenization leads to increased errors, particularly near control elements and the core periphery. (authors)
Core homogenization method for pebble bed reactors
International Nuclear Information System (INIS)
Kulik, V.; Sanchez, R.
2005-01-01
This work presents a core homogenization scheme for treating a stochastic pebble bed loading in pebble bed reactors. The reactor core is decomposed into macro-domains that contain several pebble types characterized by different degrees of burnup. A stochastic description is introduced to account for pebble-to-pebble and pebble-to-helium interactions within a macro-domain as well as for interactions between macro-domains. Performance of the proposed method is tested for the PROTEUS and ASTRA critical reactor facilities. Numerical simulations accomplished with the APOLLO2 transport lattice code show good agreement with the experimental data for the PROTEUS reactor facility and with the TRIPOLI4 Monte Carlo simulations for the ASTRA reactor configuration. The difference between the proposed method and the traditional volume-averaged homogenization technique is negligible while only one type of fuel pebbles present in the system, but it grows rapidly with the level of pebble heterogeneity. (authors)
Aqueous homogeneous suspension reactor project
International Nuclear Information System (INIS)
Kersten, J.A.H.
1976-11-01
During the period April 1 through September 30, 1976, the energy production of the KSTR reactor was rather small due to problems with the inventory determinations and a decrease of the critical temperature at power operation for which no explanation could be found. A study program in cooperation with the IAEA will be carried out on the status and prospects of thermal breeders. Work done on the investigation of KSTR samples is mentioned briefly
Homogeneous Thorium Fuel Cycles in Candu Reactors
Energy Technology Data Exchange (ETDEWEB)
Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)
2009-06-15
The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy
International Nuclear Information System (INIS)
Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.
2014-01-01
Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are
Energy Technology Data Exchange (ETDEWEB)
Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)
2014-07-01
Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the
Homogenous reactor, elaborations, not released up to end
International Nuclear Information System (INIS)
Takibayev, Zh.S.
2002-01-01
Nowadays the nuclear power uses mainly water moderated reactors, where water or heavy water works as neutron inhibitor or coolant, and fuel solid state is situated in reactor core discretely as fuel element packed in fuel assembly. Such fuel composition in solid state reactors leads to rise in price of reactor itself and, of course, many other inconveniences. Firstly, burning out depth is limited; secondary, agents absorbed neutrons are accumulated in fission products, i. e. it leads to poisoning slag derive and thirdly, there are too many outside agents in reactor core in the form of fuel elements and different constructional materials. It worsens neutron balance of reactor. There are many other inconveniences. Specialists understand this problem. They are looking for escaping of difficulty proposing to begin a wide-ranging design, for example, of a new generation of homogeneous reactor especially with salt liquid, liquid metal fuel. But this problem nowadays can not be nearly decided. It is clear enough that within at least 50-100 years the existing monopoly will not change its attitude to use of new elaboration, for example, reactor with salt liquid fuel unless a sharp necessity of opening up not only 1-2 % of uranium in the case of reactors on thermal neutrons or nearby 10-20 % for fast reactors as nowadays but effective use of all potential of nuclear fission energy contained in natural uranium and thorium resources will be realized. In the report the scheme of nuclear reactor with liquid metal or salt liquid is shown. Such approach can be in future one of possible variants of problem solution in effective opening up of all uranium-plutonium energy resource of our planet. The scheme shows only possible allocations of the container and the pipeline. Their proportioning is one of main problems of future elaborations. A mutual allocation of the container and pipelines was carried out in such way, that demand to the last ones where less than to the container
Homogeneous SLOWPOKE reactor for the production of radio-isotope. A feasibility study
Energy Technology Data Exchange (ETDEWEB)
Busatta, P.; Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)]. E-mail: paul.busatta@rmc.ca; bonin-h@rmc.ca
2006-07-01
The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP Monte Carlo reactor calculation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether natural convection can still effectively cool the reactor using the modeling software FEMLAB. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)
Homogeneous SLOWPOKE reactor for the production of radio-isotope. A feasibility study
International Nuclear Information System (INIS)
Busatta, P.; Bonin, H.W.
2006-01-01
The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP Monte Carlo reactor calculation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether natural convection can still effectively cool the reactor using the modeling software FEMLAB. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)
Homogeneous slowpoke reactor for the production of radio-isotope. A feasibility study
International Nuclear Information System (INIS)
Busatta, P.; Bonin, H.
2005-01-01
The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP 5 simulation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether the natural convection will still effectively cool the reactor using the modeling software FEMLAB. The MCNP 5 simulation code was validated by using a simulation with WIMS-AECL code. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)
Homogeneous Slowpoke reactor for the production of radio-isotope: a feasibility study
Energy Technology Data Exchange (ETDEWEB)
Busetta, P.; Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)
2006-09-15
The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP Monte Carlo reactor calculation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous react will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether natural convection can still effectively cool the reactor using the modeling software FEMLAB(r). It was found that it is needed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)
Homogeneous slowpoke reactor for the production of radio-isotope. A feasibility study
Energy Technology Data Exchange (ETDEWEB)
Busatta, P.; Bonin, H. [Royal Military College of Canada, Kingston, Ontario (Canada)]. E-mail: paul.busatta@rmc.ca; bonin-h@rmc.ca
2005-07-01
The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP 5 simulation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether the natural convection will still effectively cool the reactor using the modeling software FEMLAB. The MCNP 5 simulation code was validated by using a simulation with WIMS-AECL code. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)
The aqueous homogeneous suspension reactor project
International Nuclear Information System (INIS)
1975-01-01
The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described
Precipitation of plutonium oxalate from homogeneous solutions
International Nuclear Information System (INIS)
Rao, V.K.; Pius, I.C.; Subbarao, M.; Chinnusamy, A.; Natarajan, P.R.
1986-01-01
A method for the precipitation of plutonium(IV) oxalate from homogeneous solutions using diethyl oxalate is reported. The precipitate obtained is crystalline and easily filterable with yields in the range of 92-98% for precipitations involving a few mg to g quantities of plutonium. Decontamination factors for common impurities such as U(VI), Am(III) and Fe(III) were determined. TGA and chemical analysis of the compound indicate its composition as Pu(Csub(2)Osub(4))sub(2).6Hsub(2)O. Data are obtained on the solubility of the oxalate in nitric acid and in mixtures of nitric acid and oxalic acid of varying concentrations. Green PuOsub(2) obtained by calcination of the oxalate has specifications within the recommended values for trace foreign substances such as chlorine, fluorine, carbon and nitrogen. (author)
The aqueous homogeneous suspension reactor project
International Nuclear Information System (INIS)
1976-01-01
During 1975, reactor power has been increased to the design power of 1000 KW, whereas power fluctuations show a decrease with increased mean power. Operation experience with the reactor and associated instrumentation during 1975 is described. The results of the experiments done for fuel irradiations and investigations in the KSTR fuel, mainly to determine the amount of erosion products on the fuel, are described. Concerning the safety of operation of the KSTR reactor, several actions had to be taken mainly to replace or repair components or instrumentation of the reactor. Radiological safety and radioactivity discharges during 1975 are reported
A critical review of homogenization techniques in reactor lattices
International Nuclear Information System (INIS)
Benoist, P.
1983-01-01
The determination of the shape of the neutron flux in a whole reactor is, at the time being, a much too complex problem to be treated by transport theory. Since the earlier times of reactor theory, the necessity appeared to solve the problem in two steps. First the reactor is divided into zones, each of them forming a regular lattice. In each of these zones, homogenized parameters are determined by transport theory, in order to define an equivalent smeared medium. In a second step, these parameters are introduced in a diffusion theory scheme in order to treat the reactor as a whole. This is the homogenization procedure. 14 refs
A critical review of homogenization techniques in reactor lattices
International Nuclear Information System (INIS)
Benoist, P.
1983-01-01
The determination of the shape of the neutron flux in a whole reactor is, at the time being, a much too complex problem to be treated by transport theory. Since the earlier times of reactor theory, the necessity appeared to solve the problem in two steps. First the reactor is divided into zones, each of them forming a regular lattice. In each of these zones, homogenized parameters are determined by transport theory, in order to define an equivalent smeared medium. In a second step, these parameters are introduced in a diffusion theory scheme in order to treat the reactor as a whole. This is the homogenization procedure
Burn-up function of fuel management code for aqueous homogeneous reactors and its validation
International Nuclear Information System (INIS)
Wang Liangzi; Yao Dong; Wang Kan
2011-01-01
Fuel Management Code for Aqueous Homogeneous Reactors (FMCAHR) is developed based on the Monte Carlo transport method, to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment, searching for critical rod heights, thermal hydraulic parameters calculation, radiolytic-gas bubbles' calculation and bum-up calculation. This paper introduces the theory model and scheme of its burn-up function, and then compares its calculation results with benchmarks and with DRAGON's burn-up results, which confirms its bum-up computing precision and its applicability in the bum-up calculation and analysis for aqueous solution reactors. (authors)
Process to produce homogenized reactor fuels
International Nuclear Information System (INIS)
Hart, P.E.; Daniel, J.L.; Brite, D.W.
1980-01-01
The fuels consist of a mixture of PuO 2 and UO 2 . In order to increase the homogeneity of mechanically mixed fuels the pellets are sintered in a hydrogen atmosphere with a sufficiently low oxygen potential. This results in a reduction of Pu +4 to Pu +3 . By the reduction process water vapor is obtained increasing the pressure within the PuO 2 particles and causing PuO 2 to be pressed into the uranium oxide structure. (DG) [de
Homogenization technique for strongly heterogeneous zones in research reactors
International Nuclear Information System (INIS)
Lee, J.T.; Lee, B.H.; Cho, N.Z.; Oh, S.K.
1991-01-01
This paper reports on an iterative homogenization method using transport theory in a one-dimensional cylindrical cell model developed to improve the homogenized cross sections fro strongly heterogeneous zones in research reactors. The flux-weighting homogenized cross sections are modified by a correction factor, the cell flux ratio under an albedo boundary condition. The albedo at the cell boundary is iteratively determined to reflect the geometry effects of the material properties of the adjacent cells. This method has been tested with a simplified core model of the Korea Multipurpose Research Reactor. The results demonstrate that the reaction rates of an off-center control shroud cell, the multiplication factor, and the power distribution of the reactor core are close to those of the fine-mesh heterogeneous transport model
Directory of Open Access Journals (Sweden)
A. Isnaeni
2014-04-01
Full Text Available 99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U in the nuclear reactor. 99Mo mostly results from the fission reaction of 235U targets with a fission yield about 6.1%. A small additional amount is created from 98Mo neutron activation. Actually 99Mo is also created in the reactor fuel, but usually we do not extract it. The fuel will become spent fuel which is a highly radioactive waste. 99Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99Mo can be extracted from the fuel solution. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in an aqueous homogeneous reactor where target and fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown. With an extraction flow rate of 3.6 L/h, after 43 hours of reactor operation the production of 99Mo is relatively constant at about 98.6 curie/hour
International Nuclear Information System (INIS)
Moutsopoulos, George
2013-01-01
We solve the equations of topologically massive gravity (TMG) with a potentially non-vanishing cosmological constant for homogeneous metrics without isotropy. We only reproduce known solutions. We also discuss their homogeneous deformations, possibly with isotropy. We show that de Sitter space and hyperbolic space cannot be infinitesimally homogeneously deformed in TMG. We clarify some of their Segre–Petrov types and discuss the warped de Sitter spacetime. (paper)
Moutsopoulos, George
2013-06-01
We solve the equations of topologically massive gravity (TMG) with a potentially non-vanishing cosmological constant for homogeneous metrics without isotropy. We only reproduce known solutions. We also discuss their homogeneous deformations, possibly with isotropy. We show that de Sitter space and hyperbolic space cannot be infinitesimally homogeneously deformed in TMG. We clarify some of their Segre-Petrov types and discuss the warped de Sitter spacetime.
Control rod homogenization in heterogeneous sodium-cooled fast reactors
International Nuclear Information System (INIS)
Andersson, Mikael
2016-01-01
The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte
Mo-99 production on a LEU solution reactor
International Nuclear Information System (INIS)
Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.
2005-01-01
A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)
A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor
International Nuclear Information System (INIS)
Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.
2015-01-01
In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe
A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor
Energy Technology Data Exchange (ETDEWEB)
Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Royal Canadian Navy, Ottawa, Ontario (Canada)
2015-03-15
In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe
Revisiting homogeneous suspension reactors for production of radioisotopes
International Nuclear Information System (INIS)
Pasqualini, E.E.
2010-01-01
Some 50 years ago in Geneva Conferences I, II and III (1955. 1958 and 1964) on the Peaceful Uses of Atomic Energy, and also in Vienna Symposium on Reactor Experiments (1961), several papers where presented by different countries referring to advances in homogeneous suspension reactors. In particular the Dutch KEMA Suspension Test Reactor (KSTR) was developed, built and successfully operated in the sixties and seventies. It was a 1MWth reactor in which a suspension (6 microns spheres) of mixed UO 2 /ThO 2 in light water was circulated in a closed loop through a sphere-shaped vessel. One of the basic ideas on these suspension reactors was to apply the fission recoil separation effect as a means of purification of the fuel: the non-volatile fission products can be adsorbed in dispersed active charcoal and removed from the liquid. Undoubtedly, this method can present some advantages and better yields for the production of Mo-99 and other short lived radioisotopes, since they have to be extracted from a liquid in which practically no uranium is present. Details are mentioned of the different aspects that have been taken into account and which ones could be added in the corresponding actualization of suspension reactors for radioisotope production. In recent years great advances have been made in nanotechnology that can be used in the tailoring of fuel particles and adsorbent media. Recently, in CNEA Buenos Aires, a new facility has been inaugurated and is being equipped and licensed for laboratory experiments and preparative synthesis of nuclear nanoparticles. RA-6 and RA-3 experimental reactors in Argentina can be used for in-pile testing. (author)
Directory of Open Access Journals (Sweden)
Luo Li-Qin
2016-01-01
Full Text Available In this paper, we investigate the value distribution of meromorphic solutions of homogeneous and non-homogeneous complex linear differential-difference equations, and obtain the results on the relations between the order of the solutions and the convergence exponents of the zeros, poles, a-points and small function value points of the solutions, which show the relations in the case of non-homogeneous equations are sharper than the ones in the case of homogeneous equations.
Japanese Fast Reactor Program for Homogeneous Actinide Recycling
International Nuclear Information System (INIS)
Ishikawa, Makoto; Nagata, Takashi; Kondo, Satoru
2008-01-01
In the present report, the homogeneous actinide recycling scenario of Fast Reactor (FR) Cycle Technology Development Project (FaCT) is summarized. First, the scenario of nuclear energy policy in Japan are briefly reviewed. Second, the basic plan of Japan to manage all minor actinide (MA) by recycling is summarized objectives of which are the efficiency increase of uranium resources, the environmental burden reduction, and the increase of nuclear non-proliferation potential. Third, recent results of reactor physics study related to MA-loaded FR cores are briefly described. Fourth, typical nuclear design of MA-loaded FR cores in the FaCT project and their main features are demonstrated with the feasibility to recycle all MA in the future FR equilibrium society. Finally, the research and development program to realize the MA recycling in Japan is introduced, including international cooperation projects. (authors)
Notes on a homogeneous reactor project; Idees sur un projet de reacteur homogene
Energy Technology Data Exchange (ETDEWEB)
Benveniste, J; Bernot, J; Eidelman, D; Grenon, M; Portes, L; Raspaud, G; Tachon, J; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Berthod, L; Cohen de Lara, G; Delachanal, M; Fontanet, P; Halbronn, G [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France)
1958-07-01
An attempt has been made to develop certain ideas concerning homogeneous reactors. The project under consideration is based on the simultaneous use of a suspension of uranium dispersed in heavy or light water and of boiling in the reactor for heat extraction. However, the studies of suspensions and of boiling are relatively independent and can also be developed for reactors of different types using one or the other. Our aim is a minimum investment in fissile material; for this we propose to extract the steam directly from the core and to make use of a cyclone to accelerate this extraction; a cyclone-type circulation creating a field of increasing tangential velocities of the fluid towards the axis causes the droplets of vapour to accelerate towards the axial vortex in which they are collected; the steam output is then evacuated to the external heat utilisation system, for example an exchanger of the condenser-boiler type. The input speed of water into the reactor being one of the important parameters in the running of the pile, a spiral supply input chamber is used, allowing this speed to be regulated in amount and direction. (author)Fren. [French] Nous nous sommes attaches a developper certaines idees relatives aux piles homogenes. Le projet que nous etudions est base sur l'emploi simultane d'une suspension contenant de l'uranium disperse dans l'eau legere ou lourde et de l'ebullition dans le reacteur pour l'extraction de chaleur. Neanmoins, les etudes de suspensions et d'ebullition sont relativement independantes et peuvent egalement etre developpees pour des reacteurs de type different utilisant l'une ou l'autre. Le but que nous cherchons a atteindre est un investissement minimum en matiere fissile; pour cela, nous proposons d'extraire directement la vapeur dans le coeur et de recourir a un dispositif cyclone pour accelerer cette extraction; une circulation type cyclone creant un champ de vitesses tangentielles du fluide croissantes veraxe a pour effet d
Radiation induced homogeneous precipitation in undersaturated solid-solutions
International Nuclear Information System (INIS)
Cauvin, Richard; Martin, Georges.
1978-01-01
The stability of various types of solid solutions under irradiation is studied. In this paper, observations made on AlZn solid solutions under 1 MeV electron irradiation are reported. Al-Zn was chosen as a prototype of solid solutions with a simple miscibility gap. It is shown that under appropriate irradiation conditions undersaturated AnZn solid solutions give rise to a homogeneous precipitation of coherent G.P. zones and of incoherent Zn precipitates the atomic volume of which is smaller than that of the matrix. We propose a more general treatment of solute concentration heterogeneities in solid solutions under irradiation and suggest how it might account for the nucleation of the observed phases. The growth of the observed precipitates is studied
Coprecipitation of cadmium with copper 8-hydroxyquinolate from homogeneous solution
International Nuclear Information System (INIS)
Takiyama, Kazuyoshi; Kozen, Terumi; Ueki, Yasuyo; Ishida, Hiromi
1976-01-01
The coprecipitation of copper and cadmium 8-hydroxyquinolates from homogeneous solution was conducted from the viewpoint of crystal and analytical chemistry. To the mixed solution containing copper and cadmium ions an 8-acetoxyquinoline solution was added by keeping the pH of the solution at 9 and the resulted solution was stirred at 25 0 C. The precipitate formed at each stage of the reaction was analyzed. The precipitates in an initial stage were composed of needle crystals which characterizes copper 8-hydroxyquinolate, and were associated with a slight amount of cadmium. The first half of the coprecipitation curve for the needle crystal formation resembles the logarithmic distribution curve of lambda equal to about 0.01. The precipitation of most of the copper ions was followed by the precipitation of cadmium 8-hydroxyquinolate crystal in the plate form. The needle crystals of copper 8-hydroxyquinolate started to dissolve and transformed to plate crystals. In the second half of the coprecipitation, both crystals, owing to the identical crystal structure, precipitated simultaneously and form a solid solution. When cadmium 8-hydroxyquinolate was precipitated by the PFHS method (precipitation from homogeneous solution) in the presence of the needle crystals of copper 8-hydroxyquinolate, the above mentioned phenomenon was observed. The precipitation of cadmium 8-hydroxyquinolate in the plate form is due to the seeding effect of the plate crystals of copper 8-hydroxyquinolate, which were scantily transformed from the needle crystals. The plate crystals of cadmium compound acts as a seed to transform the needle crystals of copper compound to plate crystals. (auth.)
Fluidic delivery of homogeneous solutions through carbon tube bundles
International Nuclear Information System (INIS)
Srikar, R; Yarin, A L; Megaridis, C M
2009-01-01
A wide array of technological applications requires localized high-rate delivery of dissolved compounds (in particular, biological ones), which can be achieved by forcing the solutions or suspensions of such compounds through nano or microtubes and their bundled assemblies. Using a water-soluble compound, the fluorescent dye Rhodamine 610 chloride, frequently used as a model drug release compound, it is shown that deposit buildup on the inner walls of the delivery channels and its adverse consequences pose a severe challenge to implementing pressure-driven long-term fluidic delivery through nano and microcapillaries, even in the case of such homogeneous solutions. Pressure-driven delivery (3-6 bar) of homogeneous dye solutions through macroscopically-long (∼1 cm) carbon nano and microtubes with inner diameters in the range 100 nm-1 μm and their bundled parallel assemblies is studied experimentally and theoretically. It is shown that the flow delivery gradually shifts from fast convection-dominated (unobstructed) to slow jammed convection, and ultimately to diffusion-limited transport through a porous deposit. The jamming/clogging phenomena appear to be rather generic: they were observed in a wide concentration range for two fluorescent dyes in carbon nano and microtubes, as well as in comparable transparent glass microcapillaries. The aim of the present work is to study the physics of jamming, rather than the chemical reasons for the affinity of dye molecules to the tube walls.
Directory of Open Access Journals (Sweden)
Olaniyi Samuel Iyiola
2014-09-01
Full Text Available In this paper, we obtain analytical solutions of homogeneous time-fractional Gardner equation and non-homogeneous time-fractional models (including Buck-master equation using q-Homotopy Analysis Method (q-HAM. Our work displays the elegant nature of the application of q-HAM not only to solve homogeneous non-linear fractional differential equations but also to solve the non-homogeneous fractional differential equations. The presence of the auxiliary parameter h helps in an effective way to obtain better approximation comparable to exact solutions. The fraction-factor in this method gives it an edge over other existing analytical methods for non-linear differential equations. Comparisons are made upon the existence of exact solutions to these models. The analysis shows that our analytical solutions converge very rapidly to the exact solutions.
For the criticality of water reflected homogeneous arrays and heterogeneous reactor fuel elements
Energy Technology Data Exchange (ETDEWEB)
Mueller, Hj; Rabitsch, H; Schuerrer, F [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik
1980-01-01
The smallest critical masses for fuel elements of research reactors having a medium and high enrichment are calculated. The results fit close on the known critical masses of power reactors with low enrichment. The comparison of the critical masses of reactor fuel elements and homogenized uranium dioxide water systems yields the influence of the homogeneity and of the cladding on the criticality. A coefficient for heterogeneity is suggested which takes into consideration these influences.
Donovan, Preston; Chehreghanianzabi, Yasaman; Rathinam, Muruhan; Zustiak, Silviya Petrova
2016-01-01
The study of diffusion in macromolecular solutions is important in many biomedical applications such as separations, drug delivery, and cell encapsulation, and key for many biological processes such as protein assembly and interstitial transport. Not surprisingly, multiple models for the a-priori prediction of diffusion in macromolecular environments have been proposed. However, most models include parameters that are not readily measurable, are specific to the polymer-solute-solvent system, or are fitted and do not have a physical meaning. Here, for the first time, we develop a homogenization theory framework for the prediction of effective solute diffusivity in macromolecular environments based on physical parameters that are easily measurable and not specific to the macromolecule-solute-solvent system. Homogenization theory is useful for situations where knowledge of fine-scale parameters is used to predict bulk system behavior. As a first approximation, we focus on a model where the solute is subjected to obstructed diffusion via stationary spherical obstacles. We find that the homogenization theory results agree well with computationally more expensive Monte Carlo simulations. Moreover, the homogenization theory agrees with effective diffusivities of a solute in dilute and semi-dilute polymer solutions measured using fluorescence correlation spectroscopy. Lastly, we provide a mathematical formula for the effective diffusivity in terms of a non-dimensional and easily measurable geometric system parameter.
Directory of Open Access Journals (Sweden)
Preston Donovan
Full Text Available The study of diffusion in macromolecular solutions is important in many biomedical applications such as separations, drug delivery, and cell encapsulation, and key for many biological processes such as protein assembly and interstitial transport. Not surprisingly, multiple models for the a-priori prediction of diffusion in macromolecular environments have been proposed. However, most models include parameters that are not readily measurable, are specific to the polymer-solute-solvent system, or are fitted and do not have a physical meaning. Here, for the first time, we develop a homogenization theory framework for the prediction of effective solute diffusivity in macromolecular environments based on physical parameters that are easily measurable and not specific to the macromolecule-solute-solvent system. Homogenization theory is useful for situations where knowledge of fine-scale parameters is used to predict bulk system behavior. As a first approximation, we focus on a model where the solute is subjected to obstructed diffusion via stationary spherical obstacles. We find that the homogenization theory results agree well with computationally more expensive Monte Carlo simulations. Moreover, the homogenization theory agrees with effective diffusivities of a solute in dilute and semi-dilute polymer solutions measured using fluorescence correlation spectroscopy. Lastly, we provide a mathematical formula for the effective diffusivity in terms of a non-dimensional and easily measurable geometric system parameter.
Homogeneous Charge Compression Ignition Combustion: Challenges and Proposed Solutions
Directory of Open Access Journals (Sweden)
Mohammad Izadi Najafabadi
2013-01-01
Full Text Available Engine and car manufacturers are experiencing the demand concerning fuel efficiency and low emissions from both consumers and governments. Homogeneous charge compression ignition (HCCI is an alternative combustion technology that is cleaner and more efficient than the other types of combustion. Although the thermal efficiency and NOx emission of HCCI engine are greater in comparison with traditional engines, HCCI combustion has several main difficulties such as controlling of ignition timing, limited power output, and weak cold-start capability. In this study a literature review on HCCI engine has been performed and HCCI challenges and proposed solutions have been investigated from the point view of Ignition Timing that is the main problem of this engine. HCCI challenges are investigated by many IC engine researchers during the last decade, but practical solutions have not been presented for a fully HCCI engine. Some of the solutions are slow response time and some of them are technically difficult to implement. So it seems that fully HCCI engine needs more investigation to meet its mass-production and the future research and application should be considered as part of an effort to achieve low-temperature combustion in a wide range of operating conditions in an IC engine.
International Nuclear Information System (INIS)
Mirzaev, B.B.; Khidirov, I.; Mukhtarova, N.N.
2005-01-01
In the work by the neutron-graph the homogeneity domain of the introduction solid solution TiC x H y is determined. The sample neutron grams have been taken on the neutron diffractometer (λ=.1085 nm) installed at the thermal column of the WWR-SM reactor (INF AN RUz). For the phase analysis and estimation of solid solutions homogeneity the X-ray graph was used. X-ray grams were taken on the X-ray diffractometer DRON-3M with use of CuK α radiation (λ=0.015418 nm)
Aqueous homogeneous suspension reactor project. Report over the year 1975
Energy Technology Data Exchange (ETDEWEB)
1976-11-01
During 1975, reactor power has been increased to the design power of 1000 KW, whereas power fluctuations show a decrease with increased mean power. Operation experience with the reactor and associated instrumentation during 1975 is described. The results of the experiments done for fuel irradiations and investigations in the KSTR fuel, mainly to determine the amount of erosion products on the fuel, are described. Concerning the safety of operation of the KSTR reactor, several actions had to be taken mainly to replace or repair components or instrumentation of the reactor. Radiological safety and radioactivity discharges during 1975 are reported.
Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor
Energy Technology Data Exchange (ETDEWEB)
EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science
2015-03-15
The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.
SuPer-Homogenization (SPH) Corrected Cross Section Generation for High Temperature Reactor
Energy Technology Data Exchange (ETDEWEB)
Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hiruta, Hikaru [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2017-03-01
The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy of full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.
International Nuclear Information System (INIS)
Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco
2017-01-01
The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)
Energy Technology Data Exchange (ETDEWEB)
Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez, E-mail: milianperez89@gmail.com, E-mail: dmilian@instec.cu, E-mail: lorenapilar1109@gmail.com, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Havana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear
2017-11-01
The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly {sup 99}Mo. Compare to multipurpose research reactors, an AHR dedicated for {sup 99}Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)
International Nuclear Information System (INIS)
Yulianti, Yanti; Su'ud, Zaki; Waris, Abdul; Khotimah, S. N.; Shafii, M. Ali
2010-01-01
The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.
Energy Technology Data Exchange (ETDEWEB)
Berger, F; Bertrand, J
1958-12-01
Reactivity depends strongly on disturbances of the level of the plutonium solution In the homogeneous reactor. Proserpine has a small cylindrical core, 250 mm diameter, and 10 liters volume. With a view to reducing the dangers due to corrosion and contamination, the solution level in the core is raised by pneumatic pressure. The level is stabilized by means of a regulating system. During critical experiments the variations of the level are less than one hundredth part of a millimeter. (author) [French] Les variations du niveau de la solution de plutonium dans le reacteur homogene Proserpine ont une grosse influence sur la reactivite, car le coeur est petit (10 litres de solution dans un cylindre de diametre 250 mm). En vue de reduire les dangers dus a la corrosion et a la contamination, la commande du volume liquide est pneumatique. Nous avons realise la stabilite du niveau par une regulation qui, dans les essais en regime critique, limite les variations du plan liquide a une fraction de centieme de millimetre. (auteur)
Analysis of three idealized reactor configurations: plate, pin, and homogeneous
International Nuclear Information System (INIS)
McKnight, R.D.
1983-01-01
Detailed Monte Carlo calculations have been performed for three distinct configurations of an idealized fast critical assembly. This idealized assembly was based on the LMFBR benchmark critical assembly ZPR-6/7. In the first configuration, the entire core was loaded with the plate unit cell of ZPR-6/7. In the second configuration, the entire core was loaded with the ZPR sodium-filled pin calandria. The actual ZPR pin calandria are loaded with mixed (U,Pu) oxide pins which closely match the composition of the ZPR-6/7 plate unit cell. For the present study, slight adjustments were made in the atom concentrations and the length of the pin calandria in order to make the core boundaries and average composition for the pin-cell configuration identical to those of the plate-cell configuration. In the third configuration, the core was homogeneous, again with identical core boundaries and average composition as the plate and pin configurations
International Nuclear Information System (INIS)
Dorning, J.J.
1991-01-01
A simultaneous pin lattice cell and fuel bundle homogenization theory has been developed for use with nodal diffusion calculations of practical reactors. The theoretical development of the homogenization theory, which is based on multiple-scales asymptotic expansion methods carried out through fourth order in a small parameter, starts from the transport equation and systematically yields: a cell-homogenized bundled diffusion equation with self-consistent expressions for the cell-homogenized cross sections and diffusion tensor elements; and a bundle-homogenized global reactor diffusion equation with self-consistent expressions for the bundle-homogenized cross sections and diffusion tensor elements. The continuity of the angular flux at cell and bundle interfaces also systematically yields jump conditions for the scaler flux or so-called flux discontinuity factors on the cell and bundle interfaces in terms of the two adjacent cell or bundle eigenfunctions. The expressions required for the reconstruction of the angular flux or the 'de-homogenization' theory were obtained as an integral part of the development; hence the leading order transport theory angular flux is easily reconstructed throughout the reactor including the regions in the interior of the fuel bundles or computational nodes and in the interiors of the pin lattice cells. The theoretical development shows that the exact transport theory angular flux is obtained to first order from the whole-reactor nodal diffusion calculations, done using the homogenized nuclear data and discontinuity factors, is a product of three computed quantities: a ''cell shape function''; a ''bundle shape function''; and a ''global shape function''. 10 refs
Energy Technology Data Exchange (ETDEWEB)
Zhao, Haiqiang; Qi, Weihong, E-mail: qiwh216@csu.edu.cn; Ji, Wenhai; Wang, Tianran; Peng, Hongcheng; Wang, Qi; Jia, Yanlin; He, Jieting [Central South University, School of Materials Science and Engineering (China)
2017-05-15
Fivefold symmetry appears only in small particles and quasicrystals because internal stress in the particles increases with the particle size. However, a typical Marks decahedron with five re-entrant grooves located at the ends of the twin boundaries can further reduce the strain energy. During hydrothermal synthesis, it is difficult to stir the reaction solution contained in a digestion high-pressure tank because of the relatively small size and high-temperature and high-pressure sealed environment. In this work, we optimized a hydrothermal reaction system by replacing the conventional drying oven with a homogeneous reactor to shift the original static reaction solution into a full mixing state. Large Marks-decahedral Pd nanoparticles (~90 nm) have been successfully synthesized in the optimized hydrothermal synthesis system. Additionally, in the products, round Marks-decahedral Pd particles were also found for the first time. While it remains a challenge to understand the growth mechanism of the fivefold twinned structure, we proposed a plausible growth-mediated mechanism for Marks-decahedral Pd nanoparticles based on observations of the synthesis process.
International Nuclear Information System (INIS)
Zhao, Haiqiang; Qi, Weihong; Ji, Wenhai; Wang, Tianran; Peng, Hongcheng; Wang, Qi; Jia, Yanlin; He, Jieting
2017-01-01
Fivefold symmetry appears only in small particles and quasicrystals because internal stress in the particles increases with the particle size. However, a typical Marks decahedron with five re-entrant grooves located at the ends of the twin boundaries can further reduce the strain energy. During hydrothermal synthesis, it is difficult to stir the reaction solution contained in a digestion high-pressure tank because of the relatively small size and high-temperature and high-pressure sealed environment. In this work, we optimized a hydrothermal reaction system by replacing the conventional drying oven with a homogeneous reactor to shift the original static reaction solution into a full mixing state. Large Marks-decahedral Pd nanoparticles (~90 nm) have been successfully synthesized in the optimized hydrothermal synthesis system. Additionally, in the products, round Marks-decahedral Pd particles were also found for the first time. While it remains a challenge to understand the growth mechanism of the fivefold twinned structure, we proposed a plausible growth-mediated mechanism for Marks-decahedral Pd nanoparticles based on observations of the synthesis process.
Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation
Directory of Open Access Journals (Sweden)
Subagjo
2014-07-01
Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.
Analytical investigation of a one-dimensional homogenized model for a pressurized water reactor core
International Nuclear Information System (INIS)
Benner, J.; Schumann, U.
1981-01-01
A one-dimensional homogenized model for dynamic fluid-structure interaction in a pressurized water reactor core is used to study the influence of the virtual density and spacer's stiffness. The model consists of a linear system of partial differential equations for fluid velocity, rod velocity and pressure. For these equations analytical solutions are deduced for boundary conditions prescribing either periodic wall oscillations or linearly growing wall accelerations from rest. The theoretical model for the virtual density is verified by comparison to an experiment. For zero spacer stiffness, purely acoustic oscillations appear. For positive spacer stiffness, additional oscillations arise with relative rod motions. The wavelengths of the latter oscillations are small for weak spacers. Large numerical effort would be required in a more complete three-dimensional core-model to resolve such short wave lengths. In fact in a typical core the spacer's stiffness csub(S) is small in comparison to the fluid bulk modulus K. For csub(s)/K <= 0.1 it might be appropriate to neglect the influence of the spacers. (orig.)
Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores
International Nuclear Information System (INIS)
El Ganaoui, K.
2006-09-01
In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)
Radiation induced homogeneous precipitation in undersaturated solid-solutions
International Nuclear Information System (INIS)
Cauvin, R.; Martin, G.
1979-01-01
A TEM study of 1 MeV electron irradiated Al 1.9 at% Zn solid solution shows that Zn precipitates form, under irradiation at temperatures well above the Zn solvus temperature outside irradiation. The corresponding upward shift of this temperature is dose rate dependent. This new example of radiation-induced precipitation exhibits unexpected features, which are not accounted for by the available models: (1) no correlation exists between the location of the precipitates and that of the point defects sinks; (2) the precipitation of incoherent β-phase with atomic volume smaller than that of the matrix, and of coherent G.P. zones both occurs; (3) the size of the coherent β precipitates saturates at large dose. A general mechanism for solute concentration fluctuations under irradiation is proposed which qualitatively accounts for the formation of coherent G.P. zones and for the nucleation of solute clusters with more complex structures. A reanalysis of Russell's model (1977) for the growth of incoherent precipitates shows that it may qualitatively account for the observed behavior of the β phase precipitates. (Auth.)
Fractal solutions of recirculation tubular chemical reactors
International Nuclear Information System (INIS)
Berezowski, Marek
2003-01-01
Three kinds of fractal solutions of model of recirculation non-adiabatic tubular chemical reactors are presented. The first kind concerns the structure of Feigenbaum's diagram on the limit of chaos. The second kind and the third one concern the effect of initial conditions on the dynamic solutions of models. In the course of computations two types of recirculation were considered, viz. the recirculation of mass (return of a part of products' stream) and recirculation of heat (heat exchange in the external heat exchanger)
International Nuclear Information System (INIS)
Vanhanen, R.
2015-01-01
The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life. We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated. The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses. We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of 16 O is problematic due to lack of correlation between total and elastic reactions
Energy Technology Data Exchange (ETDEWEB)
Vanhanen, R., E-mail: risto.vanhanen@aalto.fi
2015-03-15
The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life. We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated. The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses. We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of {sup 16}O is problematic due to lack of correlation between total and elastic reactions.
Production of molybdenum-99 by heterogeneous and homogeneous uranium fueled reactors
International Nuclear Information System (INIS)
Carlin, G.E.; Bonin, H.W.
2012-01-01
The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. At the forefront of the medical isotope list is molybdenum-99 and its daughter isotope technetium-99m, which encompass over 80% of radiopharmaceutical procedures. Fission of uranium-235 to produce molybdenum-99 is the most widely used method for producing this radioisotope. The heterogeneous reactor and the aqueous homogeneous reactor are looked at here with emphasis on the use of low enriched uranium as the fuel source. Methods of technetium-99m generation and its medical use are also reviewed. (author)
Production of molybdenum-99 by heterogeneous and homogeneous uranium fueled reactors
Energy Technology Data Exchange (ETDEWEB)
Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)
2012-07-01
The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. At the forefront of the medical isotope list is molybdenum-99 and its daughter isotope technetium-99m, which encompass over 80% of radiopharmaceutical procedures. Fission of uranium-235 to produce molybdenum-99 is the most widely used method for producing this radioisotope. The heterogeneous reactor and the aqueous homogeneous reactor are looked at here with emphasis on the use of low enriched uranium as the fuel source. Methods of technetium-99m generation and its medical use are also reviewed. (author)
Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors
International Nuclear Information System (INIS)
Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A.
2004-01-01
The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO 2 -UO 2 ) fuel and compare those designs with more conventional UO 2 designs.The fuel cycle analyses indicate that ThO 2 -UO 2 fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO 2 fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO 2 -UO 2 fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO 2 fuel used in light water reactors
Giant asymmetry of separation and homogenization processes in solid 3He-4He solutions
International Nuclear Information System (INIS)
Grigor'ev, V.N.; Majdanov, V.A.; Penzev, A.A.; Polev, A.V.; Rubets, S.P.; Rudavskij, Eh.Ya.; Rybalko, A.S.; Syrnikov, E.V.
2005-01-01
The kinetics of the processes of separation and homogenization of solid 3 He- 4 He solutions is compared by using the precision barometry. The experiments were made with the initial specimens of three types: weak 3 He- 4 He and 4 He- 3 He solutions and concentrated 3 He- 4 He ones. It is found that the homogenization rate at the initial stage may be more than 500 times higher that the rate of separation. This is the case for all types of the solutions studied. The appreciable rate of phase separation in the concentrated solutions where, according to the modern concepts, impurity atoms in quantum crystals should be localized, suggests that in such conditions there is a new unknown mechanism of mass-transfer, while the fast homogenization points to a nondiffusion nature of the process
Optimal Homogenization of Perfusion Flows in Microfluidic Bio-Reactors: A Numerical Study
DEFF Research Database (Denmark)
Okkels, Fridolin; Dufva, Martin; Bruus, Henrik
2011-01-01
In recent years, the interest in small-scale bio-reactors has increased dramatically. To ensure homogeneous conditions within the complete area of perfused microfluidic bio-reactors, we develop a general design of a continually feed bio-reactor with uniform perfusion flow. This is achieved...... by introducing a specific type of perfusion inlet to the reaction area. The geometry of these inlets are found using the methods of topology optimization and shape optimization. The results are compared with two different analytic models, from which a general parametric description of the design is obtained...... and tested numerically. Such a parametric description will generally be beneficial for the design of a broad range of microfluidic bioreactors used for, e. g., cell culturing and analysis and in feeding bio-arrays....
Homogeneous SLOWPOKE reactors for Mo-99/Tc-99m production in North America
Energy Technology Data Exchange (ETDEWEB)
Hilborn, J.W., E-mail: hilbovanw@sympatico.ca [Deep River, Ontario (Canada); Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)
2014-07-01
The 15 month shutdown of NRU in 2009 - 2010 caused an overall isotope shortage of approximately 30%; and in North America, the annual Tc-99m demand decreased from an estimated 20 million unit doses to about 15 million unit doses. Mo-99/Tc-99m is produced from HEU targets, irradiated in NRU for 11 days, and after chemical removal of uranium it is shipped to Nordion in Kanata, Ontario. Nordion further purifies the material and sends it to Lantheus Medical Imaging in the USA for manufacture of Mo-99 generators, which are then distributed to hundreds of hospital radiopharmacies throughout North America. One other American company, Covidien, manufactures and distributes Mo-99 generators like Lantheus, but they import bulk Mo-99 from Europe or South Africa. At the hospitals, Tc-99m is chemically extracted daily from the Mo-99 generators and loaded into syringes for immediate clinical use. Fortuitously, the 66 hour half-life of Mo-99 allows the replenishment of Tc-99m in the generator over a growth period of about 20 hours; and a generator can be 'milked' daily for up to two weeks. A more efficient model is the direct production and distribution of Tc-99m unit doses to regional hospitals from 10 'industrial' radiopharmacies located at existing licensed reactor sites in North America. A 20 kW homogeneous SLOWPOKE reactor at each site would deliver 15 litres of irradiated uranyl sulphate fuel solution daily to industrial-scale hot cells for extraction of Mo-99, which would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and the Low Enriched Uranium (LEU) would be recycled. Each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily, for courier delivery to all of the Nuclear Medicine hospitals within a 3 hour average range by road transport. Typically, the delivered doses would be in the range 10 to 30 mCi. Assuming an average unit dose of 25 mCi at the hospital and 5 x 52
U-233 fuelled low critical mass solution reactor experiment PURNIMA II
International Nuclear Information System (INIS)
Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.
1987-01-01
A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)
Energy Technology Data Exchange (ETDEWEB)
Ithurralde, M F; Kremser, J; Leclerc, J; Lombard, Ch; Moreau, J; Robin, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1964-07-01
Criticality experiments on solutions of fissionable materials have been carried out in tanks of various geometries (cylinder, isolated annular cylinder, interacting annular cylinders); the reflexion conditions have also been varied (without reflection, semi-reflection and total reflexion by water). The range of the studied concentrations is rather large (18,8 to 104 gms/liter). The interpretation of these experiments has been undertaken in order to resolve the problems of the industrial use of homogeneous plutonium and uranium solutions. Several methods the fields of application of which are different have been used: diffusion method, transport method and Monte-Carlo method. (authors) [French] Des experiences critiques sur des solutions de matieres fissiles ont ete faites dans des cuves de diverses geometries (cylindre, cylindre annulaire isole, cylindre annulaire en interaction), les conditions de reflexion ont ete egalement variees (sans reflexion, semi reflexion et reflexion totale par l'eau). La gamme des concentrations etudiees est assez etendue (18,8 a 104 g/l ). L'interpretation de ces experiences a ete entreprise dans le but de pouvoir resoudre les problemes poses par l'emploi industriel de solutions homogenes de plutonium et d'uranium, plusieurs methodes dont les domaines d'application sont differents ont ete employees: methode de diffusion, methode de transport, methode de Monte-Carlo. (auteurs)
Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident
Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos
2012-06-01
The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High
Position-dependency of Fuel Pin Homogenization in a Pressurized Water Reactor
Energy Technology Data Exchange (ETDEWEB)
Heo, Woong; Kim, Yonghee [Korea Advanced Institute of Science and Technolgy, Daejeon (Korea, Republic of)
2016-05-15
By considering the multi-physics effects more comprehensively, it is possible to acquire precise local parameters which can result in a more accurate core design and safety assessment. A conventional approach of the multi-physics neutronics calculation for the pressurized water reactor (PWR) is to apply nodal methods. Since the nodal methods are basically based on the use of assembly-wise homogenized parameters, additional pin power reconstruction processes are necessary to obtain local power information. In the past, pin-by-pin core calculation was impractical due to the limited computational hardware capability. With the rapid advancement of computer technology, it is now perhaps quite practical to perform the direct pin-by-pin core calculation. As such, fully heterogeneous transport solvers based on both stochastic and deterministic methods have been developed for the acquisition of exact local parameters. However, the 3-D transport reactor analysis is still challenging because of the very high computational requirement. Position-dependency of the fuel pin homogenized cross sections in a small PWR core has been quantified via comparison of infinite FA and 2-D whole core calculations with the use of high-fidelity MC simulations. It is found that the pin environmental affect is especially obvious in FAs bordering the baffle reflector regions. It is also noted that the downscattering cross section is rather sensitive to the spectrum changes of the pins. It is expected that the pinwise homogenized cross sections need to be corrected somehow for accurate pin-by-pin core calculations in the peripheral region of the reactor core.
Monitoring of homogeneity of fuel compacts for high-temperature reactors
International Nuclear Information System (INIS)
Mottet, P.; Guery, M.; Chegne, J.
Apparatus using either gamma transmission or gamma scintillation spectrometry (with NaI(Tl) detector) was developed for monitoring the homogeneity of distribution of fissile and fertile particles in fuel compacts for high-temperature reactors. Three methods were studied: Longitudinal gamma transmission which gives a total distribution curve of heavy metals (U and Th); gamma spectrometry with a well type scintillator, which rapidly gives the U and Th count rates per fraction of compact; and longitudinal gamma spectrometry, giving axial distribution curves for uranium and thorium; apparatus with four scintillators and optimization of the parameters for the measurement, permitting significantly decreasing the duration of the monitoring. These relatively simple procedures should facilitate the industrial monitoring of high-temperature reactor fuel
Single-particle levitation system for automated study of homogeneous solute nucleation
Olsen, Adam P.; Flagan, Richard C.; Kornfield, Julia A.
2006-01-01
We present an instrument that addresses two critical requirements for quantitative measurements of the homogeneous crystal nucleation rate in supersaturated aqueous solution. First, the need to perform repeated measurements of nucleation incubation times is met by automating experiments to enable programmable cycling of thermodynamic conditions. Second, the need for precise and robust control of the chemical potential in supersaturated aqueous solution is met by implementing a novel technique...
DEFF Research Database (Denmark)
Brink, Bastian Klüge
work in synthesis and characterization of interstitial solutions ofnitrogen and carbon in iron-based lattices. In order to avoid the influences of gradients incomposition and residual stresses, which are typically found in treated surface layers,homogenous samples are needed. These were prepared from...
Use of LEU in the aqueous homogeneous medical isotope production reactor
Energy Technology Data Exchange (ETDEWEB)
Ball, R.M. [Babock & Wilcox, Lynchburg, VA (United States)
1997-08-01
The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.
Use of LEU in the aqueous homogeneous medical isotope production reactor
International Nuclear Information System (INIS)
Ball, R.M.
1997-01-01
The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution
An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors
International Nuclear Information System (INIS)
Tolosa, S.C.; Marajofsky, A.
2004-01-01
In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate
An expert system to analyze homogeneity in fuel element plates for research reactors
International Nuclear Information System (INIS)
Cativa Tolosa, Sebastian; Marajofsky, Adolfo
2004-01-01
In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up.This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to X-ray images. These images are generated when the X-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized X-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate. (author)
Two-dimensional analytical solution for nodal calculation of nuclear reactors
International Nuclear Information System (INIS)
Silva, Adilson C.; Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S.
2017-01-01
Highlights: • A proposal for a coarse mesh nodal method is presented. • The proposal uses the analytical solution of the two-dimensional neutrons diffusion equation. • The solution is performed homogeneous nodes with dimensions of the fuel assembly. • The solution uses four average fluxes on the node surfaces as boundary conditions. • The results show good accuracy and efficiency. - Abstract: In this paper, the two-dimensional (2D) neutron diffusion equation is analytically solved for two energy groups (2G). The spatial domain of reactor core is divided into a set of nodes with uniform nuclear parameters. To determine iteratively the multiplication factor and the neutron flux in the reactor we combine the analytical solution of the neutron diffusion equation with an iterative method known as power method. The analytical solution for different types of regions that compose the reactor is obtained, such as fuel and reflector regions. Four average fluxes in the node surfaces are used as boundary conditions for analytical solution. Discontinuity factors on the node surfaces derived from the homogenization process are applied to maintain averages reaction rates and the net current in the fuel assembly (FA). To validate the results obtained by the analytical solution a relative power density distribution in the FAs is determined from the neutron flux distribution and compared with the reference values. The results show good accuracy and efficiency.
Generation of exact solutions to the Einstein field equations for homogeneous space--time
International Nuclear Information System (INIS)
Hiromoto, R.E.
1978-01-01
A formalism is presented capable of finding all homogeneous solutions of the Einstein field equations with an arbitrary energy-stress tensor. Briefly the method involves the classification of the four-dimensional Lie algebra over the reals into nine different broad classes, using only the Lorentz group. Normally the classification of Lie algebras means that one finds all essentially different solutions of the Jacobi identities, i.e., there exists no nonsingular linear transformation which transforms two sets of structure constants into the other. This approach is to utilize the geometrical considerations of the homogeneous spacetime and field equations to be solved. Since the set of orthonormal basis vectors is not only endowed with a Minkowskian metric, but also constitutes the vector space of our four-dimensional Lie algebras, the Lie algebras are classified against the Lorentz group restricts the linear group of transformations, denoting the essentially different Lie algebras, into nine different broad classes. The classification of the four-dimensional Lie algebras represents the unification of various methods previously introduced by others. Where their methods found only specific solutions to the Einstein field equations, systematic application of the nine different classes of Lie algebras guarantees the extraction of all solutions. Therefore, the methods of others were extended, and their foundations of formalism which goes beyond the present literature of exact homogeneous solutions to the Einstein field equations is built upon
Reflection principle for classical solutions of the homogeneous real Monge–Ampère equation
Directory of Open Access Journals (Sweden)
Mika Koskenoja
2015-12-01
Full Text Available We consider reflection principle for classical solutions of the homogeneous real Monge–Ampère equation. We show that both the odd and the even reflected functions satisfy the Monge–Ampère equation if the second-order partial derivatives have continuous limits on the reflection boundary. In addition to sufficient conditions, we give some necessary conditions. Before stating the main results, we present elementary formulas for the reflected functions and study their differentiability properties across the reflection boundary. As an important special case, we finally consider extension of polynomials satisfying the homogeneous Monge–Ampère equation.
Solution of heat removal from nuclear reactors by natural convection
Directory of Open Access Journals (Sweden)
Zitek Pavel
2014-03-01
Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.
International Nuclear Information System (INIS)
Stefanovic, D.B.
1970-12-01
The objective of this work is to describe the new analytical solution of the neutron slowing down equation for infinite monoatomic media with arbitrary energy dependence of cross section. The solution is obtained by introducing Green slowing down functions instead of starting from slowing down equations directly. The previously used methods for calculation of fission neutron spectra in the reactor cell were numerical. The proposed analytical method was used for calculating the space-energy distribution of fast neutrons and number of neutron reactions in a thermal reactor cell. The role of analytical method in solving the neutron slowing down in reactor physics is to enable understating of the slowing down process and neutron transport. The obtained results could be used as standards for testing the accuracy od approximative and practical methods
MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core
Energy Technology Data Exchange (ETDEWEB)
Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)
2013-07-01
In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.
MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core
International Nuclear Information System (INIS)
Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.
2013-01-01
In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory
International Nuclear Information System (INIS)
Fenwick, John D.; Pardo-Montero, Juan
2010-01-01
Purpose: Homogenized blocked arcs are intuitively appealing as basis functions for multicriteria optimization of rotational radiotherapy. Such arcs avoid an organ-at-risk (OAR), spread dose out well over the rest-of-body (ROB), and deliver homogeneous doses to a planning target volume (PTV) using intensity modulated fluence profiles, obtainable either from closed-form solutions or iterative numerical calculations. Here, the analytic and iterative arcs are compared. Methods: Dose-distributions have been calculated for nondivergent beams, both including and excluding scatter, beam penumbra, and attenuation effects, which are left out of the derivation of the analytic arcs. The most straightforward analytic arc is created by truncating the well-known Brahme, Roos, and Lax (BRL) solution, cutting its uniform dose region down from an annulus to a smaller nonconcave region lying beyond the OAR. However, the truncation leaves behind high dose hot-spots immediately on either side of the OAR, generated by very high BRL fluence levels just beyond the OAR. These hot-spots can be eliminated using alternative analytical solutions ''C'' and ''L,'' which, respectively, deliver constant and linearly rising fluences in the gap region between the OAR and PTV (before truncation). Results: Measured in terms of PTV dose homogeneity, ROB dose-spread, and OAR avoidance, C solutions generate better arc dose-distributions than L when scatter, penumbra, and attenuation are left out of the dose modeling. Including these factors, L becomes the best analytical solution. However, the iterative approach generates better dose-distributions than any of the analytical solutions because it can account and compensate for penumbra and scatter effects. Using the analytical solutions as starting points for the iterative methodology, dose-distributions almost as good as those obtained using the conventional iterative approach can be calculated very rapidly. Conclusions: The iterative methodology is
Travelling wave solutions of the homogeneous one-dimensional FREFLO model
Huang, B.; Hong, J. Y.; Jing, G. Q.; Niu, W.; Fang, L.
2018-01-01
Presently there is quite few analytical studies in traffic flows due to the non-linearity of the governing equations. In the present paper we introduce travelling wave solutions for the homogeneous one-dimensional FREFLO model, which are expressed in the form of series and describe the procedure that vehicles/pedestrians move with a negative velocity and decelerate until rest, then accelerate inversely to positive velocities. This method is expect to be extended to more complex situations in the future.
Solution of the Lambda modes problem of a nuclear power reactor using an h–p finite element method
International Nuclear Information System (INIS)
Vidal-Ferrandiz, A.; Fayez, R.; Ginestar, D.; Verdú, G.
2014-01-01
Highlights: • An hp finite element method is proposed for the Lambda modes problem of a nuclear reactor. • Different strategies can be implemented for increasing the accuracy of the solutions. • 2D and 3D benchmarks have been studied obtaining accurate results. - Abstract: Lambda modes of a nuclear power reactor have interest in reactor physics since they have been used to develop modal methods and to study BWR reactor instabilities. An h–p-Adaptation finite element method has been implemented to compute the dominant modes the fundamental mode and the next subcritical modes of a nuclear reactor. The performance of this method has been studied in three benchmark problems, a homogeneous 2D reactor, the 2D BIBLIS reactor and the 3D IAEA reactor
International Nuclear Information System (INIS)
Hursin, Mathieu; Downar, Thomas J.; Yoon, Joo Il; Joo, Han Gyu
2016-01-01
Highlights: • Reactivity initiated accident analysis with direct whole core transient transport code. • Comparison with usual “two steps” procedure. • Effect of effective delayed neutron fraction definition on energy deposition in the fuel. • Effect of homogenized few-group cross sections generation at the assembly level on energy deposition in the fuel. • Effect of effective fuel temperature definition on energy deposition in the fuel. - Abstract: The impact of the approximations in the “two-steps” procedure used in the current generation of nodal simulators for core transient calculations is assessed by using a higher order solution obtained from a direct, whole core, transient transport calculation. A control rod ejection accident in an idealized minicore is analyzed with PARCS, which uses the two-steps procedure and DeCART which provides the higher order solution. DeCART is used as lattice code to provide the homogenized cross sections and kinetics parameters to PARCS. The approximations made by using (1) the homogenized few-group cross sections and kinetic parameters generated at the assembly level, (2) an effective delayed neutrons fraction, (3) an effective fuel temperature and (4) the few-group formulation are investigated in terms of global and local core power behavior. The results presented in the paper show that the current two-steps procedure produces sufficiently accurate transient results with respect to the direct whole core calculation solution, provided that its parameters are carefully generated using the prescriptions described in the present article.
Research on reactor physics analysis method based on Monte Carlo homogenization
International Nuclear Information System (INIS)
Ye Zhimin; Zhang Peng
2014-01-01
In order to meet the demand of nuclear energy market in the future, many new concepts of nuclear energy systems has been put forward. The traditional deterministic neutronics analysis method has been challenged in two aspects: one is the ability of generic geometry processing; the other is the multi-spectrum applicability of the multigroup cross section libraries. Due to its strong geometry modeling capability and the application of continuous energy cross section libraries, the Monte Carlo method has been widely used in reactor physics calculations, and more and more researches on Monte Carlo method has been carried out. Neutronics-thermal hydraulics coupling analysis based on Monte Carlo method has been realized. However, it still faces the problems of long computation time and slow convergence which make it not applicable to the reactor core fuel management simulations. Drawn from the deterministic core analysis method, a new two-step core analysis scheme is proposed in this work. Firstly, Monte Carlo simulations are performed for assembly, and the assembly homogenized multi-group cross sections are tallied at the same time. Secondly, the core diffusion calculations can be done with these multigroup cross sections. The new scheme can achieve high efficiency while maintain acceptable precision, so it can be used as an effective tool for the design and analysis of innovative nuclear energy systems. Numeric tests have been done in this work to verify the new scheme. (authors)
International Nuclear Information System (INIS)
Gho, C.J.
1984-10-01
Neutron transport calculation of reactors is based on the definition of homogeneized cell constants, the diffusion coefficient among others. The formalism of the evaluation of the diffusion coefficient, as also the cell model used may introduced uncertainties in results. The present study allowed to estimate these uncertainties in the case of fast neutron power reactors and criticical mockups. The validation of new simple methods and the definition of references is a consequence of this work [fr
International Nuclear Information System (INIS)
Oliveira, F.L. de; Maiorino, J.R.; Santos, R.S.
2007-01-01
This paper describes a closed form solution obtained by the expansion method for the general time dependent diffusion model with delayed emission for source transients in homogeneous media. In particular, starting from simple models, and increasing the complexity, numerical results were obtained for different types of source transients. Thus, first an analytical solution of the one group without precursors was solved, followed by considering one precursors family. The general case of G-groups with R families of precursor although having a closed form solution, cannot be solved analytically, since there are no explicit formulae for the eigenvalues, and numerical methods must be used to solve such problem. To illustrate the general solution, the multi-group (three groups) time-dependent without precursors was also solved and the results inter compared with results obtained by the previous one group models for a given fast homogeneous media, and different types of source transients. The results are being compared with the obtained by numerical methods. (author)
Degradation of aqueous phenol solutions by coaxial DBD reactor
Dojcinovic, B. P.; Manojlovic, D.; Roglic, G. M.; Obradovic, B. M.; Kuraica, M. M.; Puric, J.
2008-07-01
Solutions of 2-chlorophenol, 4-chlorophenol and 2,6-dichlorophenol in bidistilled and water from the river Danube were treated in plasma reactor. In this reactor, based on coaxial dielectric barrier discharge at atmospheric pressure, plasma is formed over a thin layer of treated water. After one pass through the reactor, starting chlorophenols concentration of 20 mg/l was diminished up to 95 %. Kinetics of the chlorophenols degradation was monitored by High Pressure Liquid Chromatography method (HPLC).
International Nuclear Information System (INIS)
Shin, Y.W.; Wiedermann, A.H.
1979-10-01
A solution method is presented for transient, homogeneous, equilibrium, two-phase flows of a single-component fluid in one space dimension. The method combines a direct finite-difference procedure and the method of characteristics. The finite-difference procedure solves the interior points of the computing domain; the boundary information is provided by a separate procedure based on the characteristics theory. The solution procedure for boundary points requires information in addition to the physical boundary conditions. This additional information is obtained by a new procedure involving integration of characteristics in the hodograph plane. Sample problems involving various combinations of basic boundary types are calculated for two-phase water/steam mixtures and single-phase nitrogen gas, and compared with independent method-of-characteristics solutions using very fine characteristic mesh. In all cases, excellent agreement is demonstrated
An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons
International Nuclear Information System (INIS)
Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.
2013-01-01
Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons
International Nuclear Information System (INIS)
Milian, D.; Milian, D. E.; Rodriguez, L. P.; Salomon, J.; Cadavid, N.
2015-01-01
99m Tc is a very useful radioisotope, which is used in nearly 80% of all nuclear medicine procedures. 99m Tc is produced from 99 Mo decay. Since 2007 the medical community has been plagued by 99 Mo shortages due to aging reactors, such as the National Research Universal reactor in Canada and the High Flux Reactor in Petten, The Netherlands. At present, most of the world's supply of 99 Mo for medical isotope production involves the neutron fission of 235 U in multipurpose research reactors. 99 Mo mostly results from the fission reaction of 235 U targets with a fission yield about 6.1%. After irradiation in the reactor, the target is digested in acid or alkaline solutions and 99 Mo is recovered through a series of extraction (separation) and purification steps. 99 Mo production system in an Aqueous Homogeneous Reactor (AHR) offers a better method, because all of the 99 Mo can be extracted from the fuel solution. Over 30 AHRs has been built and operated around the world with 149 years of combined experience. In this paper, an AHR conceptual design using LEU (Low Enriched Uranium) is optimized to meet the South American demand for 99 Mo for the coming years. Aspect related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotope production and others are evaluated. The neutronic calculations have been performed with the well-known MCNPX computational code. A benchmarking experiments performed at the Russian Research Center 'Kurchatov Institute' in order to validate that the developed models of AHRs with MCNPX code and the available library in XSDIR, ENDF/B VI.2, are adequate for studies of aqueous fuel solutions. (Author)
Solution XAS Analysis for Exploring the Active Species in Homogeneous Vanadium Complex Catalysis
Nomura, Kotohiro; Mitsudome, Takato; Tsutsumi, Ken; Yamazoe, Seiji
2018-06-01
Selected examples in V K-edge X-ray Absorption Near Edge Structure (XANES) analysis of a series of vanadium complexes containing imido ligands (possessing metal-nitrogen double bond) in toluene solution have been introduced, and their pre-edge and the edge were affected by their structures and nature of ligands. Selected results in exploring the oxidation states of the active species in ethylene dimerization/polymerization using homogeneous vanadium catalysts [consisting of (imido)vanadium(V) complexes and Al cocatalysts] by X-ray absorption spectroscopy (XAS) analyses have been introduced. It has been demonstrated that the method should provide more clear information concerning the active species in situ, especially by combination with the other methods (NMR and ESR spectra, X-ray crystallographic analysis, and reaction chemistry), and should be powerful tool for study of catalysis mechanism as well as for the structural analysis in solution.
International Nuclear Information System (INIS)
Karlsson, J.K.H.; Linden, P.
1997-01-01
The neutron transport in a bare, cylindrical and homogeneous reactor, with and without the presence of a central partially inserted control rod, has been simulated by using a Monte Carlo transport code. The behaviour of both the flux and current in this system have been investigated. We have found that the flux and especially the current are strongly affected by the presence of the control rod in its close vicinity. The results indicate the feasibility to identify the position and especially the tip of the rod from the flux and current. Further, the direction to the rod can be found from the current vector. The information content regarding the position of the rod, in both the neutron flux and the current, decays strongly as a function of distance and it is dependent on the size of the rod. In our model, the practical range over which the flux or current can be a useful indicator of the position of the tip of the rod is about 10-15 cm for a rod with a diameter of 2 cm. The practical range for identification of the position of the rod is greater for a rod of larger diameter
Solutions against PWSCC in dissimilar welds cracks of reactor components
International Nuclear Information System (INIS)
Schlader, D.; Michaut, B.; Knapp, M
2005-01-01
This article provides a brief overview of the experience accumulated by Framatome ANP in the development and use of repair and mitigation techniques of the PWSCC in dissimilar welds cracks of reactor components. A brief description of the alternatives available to the industry for the solution of this problem for both PWR and BWR reactor types is also included. These solutions have been implemented many times by Framatome ANP in Europe and the US. The article also describes the way the know-how is shared among the different regions of the company in order to offer customer specific solutions. (Author)
Directory of Open Access Journals (Sweden)
Djordjevich Alexandar
2017-12-01
Full Text Available The two-dimensional advection-diffusion equation with variable coefficients is solved by the explicit finitedifference method for the transport of solutes through a homogenous two-dimensional domain that is finite and porous. Retardation by adsorption, periodic seepage velocity, and a dispersion coefficient proportional to this velocity are permitted. The transport is from a pulse-type point source (that ceases after a period of activity. Included are the firstorder decay and zero-order production parameters proportional to the seepage velocity, and periodic boundary conditions at the origin and at the end of the domain. Results agree well with analytical solutions that were reported in the literature for special cases. It is shown that the solute concentration profile is influenced strongly by periodic velocity fluctuations. Solutions for a variety of combinations of unsteadiness of the coefficients in the advection-diffusion equation are obtainable as particular cases of the one demonstrated here. This further attests to the effectiveness of the explicit finite difference method for solving two-dimensional advection-diffusion equation with variable coefficients in finite media, which is especially important when arbitrary initial and boundary conditions are required.
Solution treatment of fast reactor claddings
International Nuclear Information System (INIS)
Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro
1974-01-01
The fuel cladding tubes for Joyo (experimental FBR) are required to be a material corresponding to AISI Type 316 and cold-rolled after solution treatment. It is necessary to have no precipitation of carbide and to make the grain size smaller than ASTM No.6. It is very difficult to obtain the fine grains without the precipitation, however. In this connection, the behavior of carbide solution at high temperature and the annealing behavior of the material cold-worked and solution-treated were studied. The following matters are described: the solid solubility line of AISI Type 316; the behavior of carbide at solution treatment temperature; and the annealing behavior of the cold-worked material. (Mori, K.)
International Nuclear Information System (INIS)
Cooling, C.M.; Williams, M.M.R.; Nygaard, E.T.; Eaton, M.D.
2013-01-01
Highlights: • A point kinetics model for the Medical Isotope Production Reactor is formulated. • Reactivity insertions are simulated using this model. • Polynomial chaos is used to simulate uncertainty in reactor parameters. • The computational efficiency of polynomial chaos is compared to that of Monte Carlo. -- Abstract: This paper models a conceptual Medical Isotope Production Reactor (MIPR) using a point kinetics model which is used to explore power excursions in the event of a reactivity insertion. The effect of uncertainty of key parameters is modelled using intrusive polynomial chaos. It is found that the system is stable against reactivity insertions and power excursions are all bounded and tend towards a new equilibrium state due to the negative feedbacks inherent in Aqueous Homogeneous Reactors (AHRs). The Polynomial Chaos Expansion (PCE) method is found to be much more computationally efficient than that of Monte Carlo simulation in this application
Logistics of the research reactor fuel cycle: AREVA solutions
International Nuclear Information System (INIS)
Ohayon, David; Halle, Laurent; Naigeon, Philippe; Falgoux, Jean-Louis; Franck Obadia, Franck; Auziere, Philippe
2005-01-01
The AREVA Group Companies offer comprehensive solutions for the entire fuel cycle of Research Reactors comply with IAEA standards. CERCA and Cogema Logistics have developed a full partnership in the front end cycle. In the field of uranium CERCA and Cogema Logistics have the long term experience of the shipment from Russia, USA to the CERCA plant.. Since 1960, CERCA has manufactured over 300,000 fuel plates and 15,000 fuel elements of more than 70 designs. These fuel elements have been delivered to 40 research reactors in 20 countries. For the Back-End stage, Cogema and Cogema Logistics propose customised solutions and services for international shipments. Cogema Logistics has developed a new generation of packaging to meet the various needs and requirements of the Laboratories and Research Reactors all over the world, and complex regulatory framework. Comprehensive assistance dedicated, services, technical studies, packaging and transport systems are provided by AREVA for every step of research reactor fuel cycle. (author)
Li, Li; Li, YanYan; Yan, Xukai
2018-03-01
We classify all (-1)-homogeneous axisymmetric no-swirl solutions of incompressible stationary Navier-Stokes equations in three dimension which are smooth on the unit sphere minus the south pole, parameterize them as a two dimensional surface with boundary, and analyze their pressure profiles near the north pole. Then we prove that there is a curve of (-1)-homogeneous axisymmetric solutions with nonzero swirl, having the same smoothness property, emanating from every point of the interior and one part of the boundary of the solution surface. Moreover we prove that there is no such curve of solutions for any point on the other part of the boundary. We also establish asymptotic expansions for every (-1)-homogeneous axisymmetric solutions in a neighborhood of the singular point on the unit sphere.
International Nuclear Information System (INIS)
Newton, T.D.
1988-01-01
This paper examines the application of homogeneous equivalent absorber rod cross-sections to the calculation of control rod anti-reactivities in large fast reactors. The method used to obtain the equivalent cross-sections is described and their validity in simple whole core geometry calculations is verified. Finally, they are employed in the calculation of control rod anti-reactivity worths in the Super Phenix 1 fast reactor and the results are compared with measured values. (author). 5 refs, 5 figs, 9 tabs
Calculation on maximum accumulation of Pu-239 and Pu-241 from aqueous homogeneous reactor
International Nuclear Information System (INIS)
Ikhlas H Siregar; Frida Agung R; Suharyana; Azizul Khakim; Dahman Siregar
2016-01-01
Calculations on maximum accumulation of Pu-239 and Pu-241 using MCNPX computer code with UO_2(NO_3)_2 fuel solution enriched by 19.75% operating at temperature 80°C have been conducted. AHR design was simulated with cylindrical core having diameter of 63.4 cm and 122 cm high. From this geometry we found the reactor was critical with density 108 gr U/L of UO_2(NO_3)_2 solution. The result showed that multiplication factor (k_e_f_f) of AHR was 1.05284. Then the burn up calculations were done for various time intervals from 5 days until 285 years to analyze the result. From calculation, it was found out that the saturated concentration of Pu-239 was reached after 40-50 years of operation, producing 1.23 x 102 gr and the activity 7.645 Ci. While for operate time of AHR to produce Pu-241 should under 80 years with mass 21.7 gr and the activity 2.247 x 103 Ci. The accumulations of both isotopes are considered to be small, having low potential for misusing them for producing nuclear weapon. (author)
DiPerna, Daniel T; Blake, William K; DiPerna, Xingguang Z
2006-12-01
A formulation is developed to predict the vibration response of a finite length, submerged plate due to a line drive. The formulation starts by describing the fluid in terms of elliptic cylinder coordinates, which allows the fluid loading term to be expressed in terms of Mathieu functions. By moving the fluid loading term to the right-hand side of the equation, it is considered to be a force. The operator that remains on the left-hand side is the same as that of the in vacuo plate: a fourth-order, constant coefficient, ordinary differential equation. Therefore, the problem appears to be an inhomogeneous ordinary differential equation. The solution that results has the same form as that of the in vacuo plate: the sum of a forced solution, and four homogeneous solutions, each of which is multiplied by an arbitrary constant. These constants are then chosen to satisfy the structural boundary conditions on the two ends of the plate. Results for the finite plate are compared to the infinite plate in both the wave number and spatial domains. The theoretical predictions of the plate velocity response are also compared to results from finite element analysis and show reasonable agreement over a large frequency range.
International Nuclear Information System (INIS)
Tran Ngoc, T.D.
2008-07-01
This Ph.D thesis presents the development of the solute transport models in unsaturated double-porosity medium, by using the asymptotic homogenization method. The obtained macroscopic models concern diffusion, diffusion-convection and dispersion-convection, according to the transport regime which is characterized by the non-dimensional numbers. The models consist of two coupled equations that show the local non-equilibrium of concentrations. The double-porosity transport models were numerically implemented using the code COMSOL Multiphysics (finite elements method), and compared with the solution of the same problem at the fine scale. The implementation allows solving the coupled equations in the macro- and micro-porosity domains (two-scale computations). The calculations of the dispersion tensor as a solution of the local boundary value problems, were also conducted. It was shown that the dispersivity depends on the saturation, the physical properties of the macro-porosity domain and the internal structure of the double-porosity medium. Finally, two series of experiments were performed on a physical model of double-porosity that is composed of a periodic assemblage of sintered clay spheres in Hostun sand HN38. The first experiment was a drainage experiment, which was conducted in order to validate the unsaturated flow model. The second series was a dispersion experiment in permanent unsaturated water flow condition (water content measured by gamma ray attenuation technique). A good agreement between the numerical simulations and the experimental observations allows the validation of the developed models. (author)
Bars, Itzhak; Chen, Shih-Hung; Steinhardt, Paul J.; Turok, Neil
2012-10-01
We study a model of a scalar field minimally coupled to gravity, with a specific potential energy for the scalar field, and include curvature and radiation as two additional parameters. Our goal is to obtain analytically the complete set of configurations of a homogeneous and isotropic universe as a function of time. This leads to a geodesically complete description of the Universe, including the passage through the cosmological singularities, at the classical level. We give all the solutions analytically without any restrictions on the parameter space of the model or initial values of the fields. We find that for generic solutions the Universe goes through a singular (zero-size) bounce by entering a period of antigravity at each big crunch and exiting from it at the following big bang. This happens cyclically again and again without violating the null-energy condition. There is a special subset of geodesically complete nongeneric solutions which perform zero-size bounces without ever entering the antigravity regime in all cycles. For these, initial values of the fields are synchronized and quantized but the parameters of the model are not restricted. There is also a subset of spatial curvature-induced solutions that have finite-size bounces in the gravity regime and never enter the antigravity phase. These exist only within a small continuous domain of parameter space without fine-tuning the initial conditions. To obtain these results, we identified 25 regions of a 6-parameter space in which the complete set of analytic solutions are explicitly obtained.
Li, Li; Li, YanYan; Yan, Xukai
2018-05-01
We classify all (- 1)-homogeneous axisymmetric no-swirl solutions of incompressible stationary Navier-Stokes equations in three dimension which are smooth on the unit sphere minus the south and north poles, parameterizing them as a four dimensional surface with boundary in appropriate function spaces. Then we establish smoothness properties of the solution surface in the four parameters. The smoothness properties will be used in a subsequent paper where we study the existence of (- 1)-homogeneous axisymmetric solutions with non-zero swirl on S2 ∖ { S , N }, emanating from the four dimensional solution surface.
Aqueous homogeneous suspension reactor project. Report over the 4th quarter and the year 1974
Energy Technology Data Exchange (ETDEWEB)
1975-07-01
The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described.
The aqueous homogeneous suspension reactor project. Report over the 2nd quarter 1975
Energy Technology Data Exchange (ETDEWEB)
1975-07-01
Operation experiences, the behaviour of the reactor at 1000 KW, and the performance of reactor instruments are reported. Due to the rise in costs of the primary materials, the costs of the uranium consumption and the fuel cycle costs of the KSTR reactor are recalculated. Experiments done during the reporting period are described. Work carried out by the Health Physics Department and work carried out in connection with reactor safety is described.
Farsiani, Yasaman; Elbing, Brian
2017-11-01
High molecular weight polymer solutions in wall-bounded flows can reduce the local skin friction by as much as 80%. External flow studies have typical focused on injection of polymer within a developing turbulent boundary layer (TBL), allowing the concentration and drag reduction level to evolve with downstream distance. Modification of the log-law region of the TBL is directly related to drag reduction, but recent results suggest that the exact behavior is dependent on flow and polymer properties. Weissenberg number and the viscosity ratio (ratio of solvent viscosity to the zero-shear viscosity) are concentration dependent, thus the current study uses a polymer ocean (i.e. a homogenous concentration of polymer solution) with a developing TBL to eliminate uncertainty related to polymer properties. The near-wall modified TBL velocity profiles are acquired with particle image velocimetry. In the current presentation the mean velocity profiles and the corresponding flow (Reynolds number) and polymer (Weissenberg number, viscosity ratio, and length ratio) properties are reported. Note that the impact of polymer degradation on molecular weight will also be quantified and accounted for when estimating polymer properties This work was supported by NSF Grant 1604978.
An innovative nuclear reactor as a solution to global warming
International Nuclear Information System (INIS)
Silva, Robson Silva da; Sefidvash, Farhang
2007-01-01
The problem of global warming is no longer a philosophical discussion, but it is a fact seriously threatening the future of humanity. In this paper a practical solution to the problem of global warming resulting from the fossil fuelled energy suppliers is presented. The energy conservation and alternative forms of energy such as solar, wind, and bio even though having important roles, do not satisfy the energy demand generated by an increasing world population that desires to increase its standard of living. The fission process in the nuclear reactors does not produce greenhouse gases that cause global warming. The new paradigm in nuclear energy is the future innovative reactors that meet the new standards set by the INPRO Program of the IAEA. One such a reactor is presented in this paper, namely the Fixed Bed Nuclear Reactor (FBNR) that is supported by the International Atomic Energy (IAEA) in its program of Small Reactors Without On-Site Refuelling (SRWOSR), being one of the four water cooled reactors in this program. The other three reactor concepts are PFPWR50 of Japan, BWRPB of Russia and AFPR-100 of USA. It is shown that the nuclear energy of the future is totally different than what is today in respect to safety, economics, environmental impact and proliferation. In this manner, the public perception of nuclear energy will change and its acceptability is promoted. (author)
International Nuclear Information System (INIS)
Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Yoon-Sang; Lee, Don-Bae; Sohn, Woong-Hee; Hong, Soon-Hyung
1998-01-01
A study on improving the homogeneous dispersion of atomized spherical particles in fuel meats has been performed in connection with the development of high uranium density fuel. In comparing various mixing methods, the better homogeneity of the mixture could be obtained as in order of Spex mill, V-shape tumbler mixer, and off-axis rotating drum mixer. The Spex mill mixer required some laborious work because of its small capacity per batch. Trough optimizing the rotating speed parameter for the V-shape tumbler mixer, almost the same homogeneity as with the Spex mill could be obtained. The homogeneity of the extruded fuel meats appeared to improve through extrusion. All extruded fuel meats with U 3 Si powder of 50-volume % had fairly smooth surfaces. The homogeneity of fuel meats by V-shaped tumbler mixer revealed to be fairly good on micrographs. (author)
LPV model development and control of a solution copolymerization reactor
Rahme, S.; Abbas, H.M.S.; Meskin, N.; Tóth, R.; Mohammadpour, J.
2016-01-01
In this paper, linear parameter-varying (LPV) control is considered for a solution copolymerization reactor, which takes into account the time-varying nature of the parameters of the process. The nonlinear model of the process is first converted to an exact LPV model representation in the
Review of Kaganove's solution for the reactor point kinetics equations
International Nuclear Information System (INIS)
Couto, R.T.; Santo, A.C.F. de.
1993-09-01
A review of Kaganove's method for the reactor point kinetics equations solution is performed. This was method chosen to calculate the power in ATR, a computer program for the analysis of reactivity transients. The reasons for this choice and the adaptation of the method to the purposes of ATR are presented. (author)
Souto Mantecon, Francisco Javier
One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite
Numerical solution for identification of feedback coefficients in nuclear reactors
International Nuclear Information System (INIS)
Ebizuka, Yoshie; Sakai, Hideo
1975-01-01
Quasilinearization technique was studied to determine the Kinetic parameters of nuclear reactors. The method of solution was generalized to the determination of the parameters contained in a nonlinear system with nonlinear boundary conditions. A computer program, SNR-3, was developed to solve the resulting nonlinear two-point boundary value equations with generalized boundary conditions. In this paper, the problem formulation and the method of solution are explained for a general type of time dependent problem. A flow chart shows the procedure of numerical solution. The method was then applied to the determination of the critical factor and the reactivity feedback coefficients of reactors to investigate the accuracy and the applicability of the present method. The results showed that the present method was considerably successful, but that the random observation error effected the results of the identification. (Aoki, K.)
Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6
International Nuclear Information System (INIS)
Liu Guisheng
1995-11-01
How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs
Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6
International Nuclear Information System (INIS)
Liu Guisheng
1995-01-01
How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)
International Nuclear Information System (INIS)
Suda, S.; Franssen, F.
1987-01-01
A safeguards verification technique is being developed for determining whether process-liquid homogeneity has been achieved in process tanks and for authenticating volume-measurement algorithms involving temperature corrections. It is proposed that, in new designs for bulk-handling plants employing automated process lines, bubbler probes and thermocouples be installed at several heights in key accountability tanks. High-accuracy measurements of density using an electromanometer can now be made which match or even exceed analytical-laboratory accuracies. Together with regional determination of tank temperatures, these measurements provide density, liquid-column weight and temperature gradients over the fill range of the tank that can be used to ascertain when the tank solution has reached equilibrium. Temperature-correction algorithms can be authenticated by comparing the volumes obtained from the several bubbler-probe liquid-height measurements, each based on different amounts of liquid above and below the probe. The verification technique is based on the automated electromanometer system developed by Brookhaven National Laboratory (BNL). The IAEA has recently approved the purchase of a stainless-steel tank equipped with multiple bubbler and thermocouple probes for installation in its Bulk Calibration Laboratory at IAEA Headquarters, Vienna. The verification technique is scheduled for preliminary trials in late 1987
Forward flux sampling calculation of homogeneous nucleation rates from aqueous NaCl solutions.
Jiang, Hao; Haji-Akbari, Amir; Debenedetti, Pablo G; Panagiotopoulos, Athanassios Z
2018-01-28
We used molecular dynamics simulations and the path sampling technique known as forward flux sampling to study homogeneous nucleation of NaCl crystals from supersaturated aqueous solutions at 298 K and 1 bar. Nucleation rates were obtained for a range of salt concentrations for the Joung-Cheatham NaCl force field combined with the Extended Simple Point Charge (SPC/E) water model. The calculated nucleation rates are significantly lower than the available experimental measurements. The estimates for the nucleation rates in this work do not rely on classical nucleation theory, but the pathways observed in the simulations suggest that the nucleation process is better described by classical nucleation theory than an alternative interpretation based on Ostwald's step rule, in contrast to some prior simulations of related models. In addition to the size of NaCl nucleus, we find that the crystallinity of a nascent cluster plays an important role in the nucleation process. Nuclei with high crystallinity were found to have higher growth probability and longer lifetimes, possibly because they are less exposed to hydration water.
Mozhaev, V V; Poltevsky, K G; Slepnev, V I; Badun, G A; Levashov, A V
1991-11-04
A typical hydrophilic enzyme, CT, can be dissolved in nonpolar organic solvents (n-octane, cyclohexane and toluene) up to microM concentrations. In the homogeneous solution obtained, the enzyme possesses catalytic activity and enormously high thermostability. It does not lose this activity even after several hours refluxing in octane (126 degrees C) or cyclohexane (81 degrees C).
Photoactive TiO2 prepared by homogenous precipitation of aqueos solution of Ti4+ salt with urea
Czech Academy of Sciences Publication Activity Database
Šubrt, Jan; Bakardjieva, Snejana; Hostomský, Jiří; Jirkovský, Jaromír; Maguela, L. A. P.; Hálová, Jaroslava
2003-01-01
Roč. 12, č. 3 (2003), s. 423-428 ISSN 1453-7672 R&D Projects: GA ČR GA203/02/0983 Institutional research plan: CEZ:AV0Z4040901; CEZ:AV0Z4032918 Keywords : homogenous precipitation * aqueous solutions Subject RIV: CA - Inorganic Chemistry
International Nuclear Information System (INIS)
Gulik, Volodymyr; Tkaczyk, Alan H.
2014-01-01
Highlights: • The optimization of two-zone homogeneous subcritical systems has been performed. • A Serpent model for two-zone heterogeneous subcritical systems has been developed. • The optimization of two-zone heterogeneous subcritical systems has been carried out. • Economically optimal core composition of two-zone subcritical system was found. • The neutron spectra of the heterogeneous subcritical systems have been obtained. - Abstract: Subcritical systems driven by external neutron sources, commonly known as Accelerator-Driven System (ADS), are one type of advanced nuclear reactor exhibiting attractive characteristics, distinguished from the traditional critical systems by their intrinsic safety features. In addition, an ADS can be used for the transmutation of the nuclear waste, accumulated during the operation of existing reactors. The optimization of a subcritical nuclear reactor in terms of materials (fuel content, coolant, etc.), geometrical, and economical parameters is a crucial step in the process of their design and construction. This article describes the optimization modeling performed for homogeneous and heterogeneous two-zone subcritical systems in terms of geometry of the fuel zones. Economical assessment was also carried out for the costs of the fuel in the core of the system. Optimization modeling was performed with the Serpent-1.1.18 Monte Carlo code. The model of a two-zone subcritical system with a fast inner and a thermal gas-cooled graphite-moderated outer zone was developed, simulated, and analyzed. The optimal value for the pitch of fuel elements in the thermal outer zone was investigated from the viewpoint of the cost of subcritical system. As the main goal of ADS development is nuclear waste transmutation, neutron spectra for both fast and thermal zones were obtained for different system configurations. The results of optimization modeling of homogeneous and heterogeneous two-zone subcritical systems show that an optimal
Directory of Open Access Journals (Sweden)
Ravi Borana
2016-09-01
Full Text Available In the petroleum reservoir at an early stage the oil is recovered due to existing natural pressure and such type of oil recovery is referred as primary oil recovery. It ends when pressure equilibrium occurs and still large amount of oil remains in the reservoir. Consequently, secondary oil recovery process is employed by injection water into some injection wells to push oil towards the production well. The instability phenomenon arises during secondary oil recovery process. When water is injected into the oil filled region, due to the force of injecting water and difference in viscosities of water and native oil, protuberances occur at the common interface. It gives rise to the shape of fingers (protuberances at common interface. The injected water shoots through inter connected capillaries at very high speed. It appears in the form of irregular trembling fingers, filled with injected water in the native oil field; this is due to the immiscibility of water and oil. The homogeneous porous medium is considered with a small inclination with the horizontal, the basic parameters porosity and permeability remain uniform throughout the porous medium. Based on the mass conservation principle and important Darcy's law under the specific standard relationships and basic assumptions considered, the governing equation yields a non-linear partial differential equation. The Crank–Nicolson finite difference scheme is developed and on implementing the boundary conditions the resulting finite difference scheme is implemented to obtain the numerical results. The numerical results are obtained by generating a MATLAB code for the saturation of water which decreases with the space variable and increases with time. The obtained numerical solution is efficient, accurate, and reliable, matches well with the physical phenomenon.
Energy Technology Data Exchange (ETDEWEB)
Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)
2013-07-01
The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. New interest has been found in the use of liquid fueled nuclear reactors to produce these isotopes due to the ease of fuel processing and ability to efficiently use LEU as the fuel source. A version of this reactor is being developed at the Royal Military College of Canada to act as a successor to the SLOWPOKE-2 platform. The thermal hydraulic and transient characteristics of a 20 kWt version are being studied to verify inherent safety abilities. (author)
Solution of the reactor point kinetics equations by MATLAB computing
Directory of Open Access Journals (Sweden)
Singh Sudhansu S.
2015-01-01
Full Text Available The numerical solution of the point kinetics equations in the presence of Newtonian temperature feedback has been a challenging issue for analyzing the reactor transients. Reactor point kinetics equations are a system of stiff ordinary differential equations which need special numerical treatments. Although a plethora of numerical intricacies have been introduced to solve the point kinetics equations over the years, some of the simple and straightforward methods still work very efficiently with extraordinary accuracy. As an example, it has been shown recently that the fundamental backward Euler finite difference algorithm with its simplicity has proven to be one of the most effective legacy methods. Complementing the back-ward Euler finite difference scheme, the present work demonstrates the application of ordinary differential equation suite available in the MATLAB software package to solve the stiff reactor point kinetics equations with Newtonian temperature feedback effects very effectively by analyzing various classic benchmark cases. Fair accuracy of the results implies the efficient application of MATLAB ordinary differential equation suite for solving the reactor point kinetics equations as an alternate method for future applications.
Energy Technology Data Exchange (ETDEWEB)
Tumelero, Fernanda; Bodmann, Bardo E. J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul (PROMEC/UFRGS), Porto Alegre, RS (Brazil). Programa de Pos Graduacao em Engenharia Mecanica; Lapa, Celso M.F., E-mail: fernanda.tumelero@yahoo.com.br, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com, E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
In this work we solve the space kinetic diffusion equation in a one-dimensional geometry considering a homogeneous domain, for two energy groups and six groups of delayed neutron precursors. The proposed methodology makes use of a Taylor expansion in the space variable of the scalar neutron flux (fast and thermal) and the concentration of delayed neutron precursors, allocating the time dependence to the coefficients. Upon truncating the Taylor series at quadratic order, one obtains a set of recursive systems of ordinary differential equations, where a modified decomposition method is applied. The coefficient matrix is split into two, one constant diagonal matrix and the second one with the remaining time dependent and off-diagonal terms. Moreover, the equation system is reorganized such that the terms containing the latter matrix are treated as source terms. Note, that the homogeneous equation system has a well known solution, since the matrix is diagonal and constant. This solution plays the role of the recursion initialization of the decomposition method. The recursion scheme is set up in a fashion where the solutions of the previous recursion steps determine the source terms of the subsequent steps. A second feature of the method is the choice of the initial and boundary conditions, which are satisfied by the recursion initialization, while from the rst recursion step onward the initial and boundary conditions are homogeneous. The recursion depth is then governed by a prescribed accuracy for the solution. (author)
International Nuclear Information System (INIS)
Benner, J.
1984-03-01
A method for the numerical simulation of the Pressurized Water Reactor (PWR) core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. All these models have been implemented into the code Flux-4. For the solution of the very complex, coupled equations of motions for fluid and fuel rods an efficient numerical solution technique has been developed. With the new code-version Flux-5 the PWR-blowdown is parametically investigated. The calculated core barrel loadings are compared with Flux-4 results, simulating the core's inertia by a mass ring of HDR type. (orig.) [de
Homogenization of the internal structures of a reactor with the cooling fluid
Energy Technology Data Exchange (ETDEWEB)
Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Bliard, F. [Socotec Industrie, Service AME, 78 - Montigny le Bretonneux (France)
2001-07-01
To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)
Homogenization of the internal structures of a reactor with the cooling fluid
International Nuclear Information System (INIS)
Robbe, M.F.; Bliard, F.
2001-01-01
To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)
Energy Technology Data Exchange (ETDEWEB)
Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Garcia, Lorena P. Rodriguez; Llanes, Jesus Salomon; Hernandez, Carlos R. Garcia, E-mail: dperez@instec.cu, E-mail: dmilian@instec.cu, E-mail: lorenapilar@instec.cu, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Lira, Carlos A. Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife (Brazil); Rodriguez, Manuel Cadavid, E-mail: mcadavid2001@yahoo.com [Tecnologia Nuclear Medica Spa, TNM (Chile)
2015-07-01
{sup 99m}Tc is the most common radioisotope used in nuclear medicine. It is a very useful radioisotope, which is used in about 30-40 million procedures worldwide every year. Medical diagnostic imaging techniques using {sup 99m}Tc represent approximately 80% of all nuclear medicine procedures. Although {sup 99m}Tc can be produced directly on a cyclotron or other type of particle accelerator, currently is almost exclusively produced from the beta-decay of its 66-h parent {sup 99}Mo. {sup 99}Mo production system in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing and purification characteristics. In this paper, an AHR conceptual design using Low Enriched Uranium (LEU) is studied and optimized for the production of {sup 99}Mo. Aspects related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotopes production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution due to thermal expansion of the fuel solution and the void volume generated by radiolytic gas bubbles were evaluated. Thermal-hydraulics studies were carried out in order to show that sufficient cooling capacity exists to prevent fuel overheating. The neutronic and thermal-hydraulics calculations have been performed with the MCNPX computational code and the version 14 of ANSYS CFX respectively. The neutronic calculations demonstrated that the reactor is able to produce 370 six-day curies of {sup 99}Mo in 5 days operation cycles and the CFD simulation demonstrated that the heat removal systems provide sufficient cooling capacity to prevent fuel overheating, the maximum temperature reached by the fuel (89.29 deg C) was smaller to the allowable temperature limit (90 deg C). (author)
Maci, S.; Neto, A.
2004-01-01
This second part of a two-paper sequence deals with the uniform asymptotic description of the Green's function of an infinite slot printed between two different homogeneous dielectric media. Starting from the magnetic current derived in Part I, the dyadic green's function is first formulated in
Physically - engineering problems of the Salaspils Nuclear reactor: Solutions and their topicality
International Nuclear Information System (INIS)
Mozgirs, Z.V.
2005-01-01
The paper generalizes technical solutions of physically-engineering problems of the Salaspils nuclear research reactor, experience of its modernization and exploitation. New equipment and the related technical solutions have been tested at the Salaspils reactor during its operation time and are now recommended for further use at nuclear reactors. (author)
International Nuclear Information System (INIS)
Chen, K.F.; Olson, C.A.
1983-01-01
One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction
A second stage homogenization method
International Nuclear Information System (INIS)
Makai, M.
1981-01-01
A second homogenization is needed before the diffusion calculation of the core of large reactors. Such a second stage homogenization is outlined here. Our starting point is the Floquet theorem for it states that the diffusion equation for a periodic core always has a particular solution of the form esup(j)sup(B)sup(x) u (x). It is pointed out that the perturbation series expansion of function u can be derived by solving eigenvalue problems and the eigenvalues serve to define homogenized cross sections. With the help of these eigenvalues a homogenized diffusion equation can be derived the solution of which is cos Bx, the macroflux. It is shown that the flux can be expressed as a series of buckling. The leading term in this series is the well known Wigner-Seitz formula. Finally three examples are given: periodic absorption, a cell with an absorber pin in the cell centre, and a cell of three regions. (orig.)
An evaluation of once-through homogeneous thorium fuel cycle for light water reactors
International Nuclear Information System (INIS)
Joo, H. K.; Noh, J. M.; Yoo, J. W.
2002-01-01
The other ways enhancing the economic potential of thorium fuel has been assessed ; the utilization of lower enriched uranium in thorium-uranium fuel, duplex thorium fuel concept, thorium utilization in the mixed core with uranium fuel assembly and thorium blanket utilization in the uranium core. The fuel economics of the proposed ways of thorium fuel increased compared to the previous homogeneous thorium fuel cycle. Compared to uranium fuel cycle, however, they do not show any economic incentives. From the view of proliferation resistance potential, thorium fuel option has the advantage to reduce the inventory of plutonium production. Any of proposed thorium options are less economical than uranium fuel option, the thorium fuel option has the potential to be utilized in the future for the sake of the effective consumption of excessive plutonium and the preparation against the using up of uranium resource
An assessment of once-through homogeneous thorium fuel economics for light water reactors
International Nuclear Information System (INIS)
Joo, Hyung Kook; Noh, Jae Man; Yoo, Jae Woon
2001-01-01
The fuel economics of an once-through homogeneous thorium fuel concept for PWR was assessed by doing a detailed core analysis. In addition to this, the fuel economics assessment was also performed for two other ways enhancing the economic potential of thorium fuel; thorium utilization in the mixed core with uranium fuel assembly and Duplex thorium fuel concepts. As a results of fuel economics assessment, the thorium fuel cycle does not show any economic incentives in preference to uranium fuel cycle under the 18-months fuel cycle for PWR. However, the utilization of thorium is the mixed core with uranium fuel assembly and Duplex thorium fuel cycle and show superior fuel economics to uranium fuel under the longer fuel cycle scheme. The economic potential of once-through thorium fuel cycle is expected to be increased further by utilizing the Duplex thorium fuel in the mixed core with uranium fuel assembly
Lam, Nghi Q.; Janghorban, K.; Ardell, A. J.
1981-10-01
Irradiation-induced solute redistribution leading to precipitation of coherent γ' particles in undersaturated Ni-based solid solutions containing 6 and 8 at.% Si during 400-keV proton bombardment was modeled, based on the concept of solute segregation in concentrated alloys under spatially-dependent defect production conditions. The combined effects of (i) an extremely large difference between the defect production rates in the peak-damage and mid-range regions during irradiation and (ii) a preferential coupling between the interstitial and solute fluxes generate a net transient flux of Si atoms into the mid-range region, which is much larger than the solute flux out of this location. As a result, the Si concentration exceeds the solubility limit and homogeneous precipitation of the γ' phase occurs in this particular region of the irradiated samples. The spatial, compositional and temperature dependences of irradiation-induced homogeneous precipitation derived from the present theoretical calculations are in good qualitative agreement with experimental observations
International Nuclear Information System (INIS)
Oliveira, Fernando Luiz de
2008-01-01
This work describes an analytical solution obtained by the expansion method for the spatial kinetics using the diffusion model with delayed emission for source transients in homogeneous media. In particular, starting from simple models, and increasing the complexity, numerical results were obtained for different types of source transients. An analytical solution of the one group without precursors was solved, followed by considering one precursors family. The general case of G-groups with R families of precursor although having a closed form solution, cannot be solved analytically, since there are no explicit formulae for the eigenvalues, and numerical methods must be used to solve such problem. To illustrate the general solution, the multi-group (three groups) time-dependent problem without precursors was solved and the numerical results of a finite difference code were compared with the exact results for different transients. (author)
Parrish, K. E.; Zhang, J.; Teasdale, E.
2007-12-01
An exact analytical solution to the ordinary one-dimensional partial differential equation is derived for transient groundwater flow in a homogeneous, confined, horizontal aquifer using Laplace transformation. The theoretical analysis is based on the assumption that the aquifer is homogeneous and one-dimensional (horizontal); confined between impermeable formations on top and bottom; and of infinite horizontal extent and constant thickness. It is also assumed that there is only a single pumping well penetrating the entire aquifer; flow is everywhere horizontal within the aquifer to the well; the well is pumping with a constant discharge rate; the well diameter is infinitesimally small; and the hydraulic head is uniform throughout the aquifer before pumping. Similar to the Theis solution, this solution is suited to determine transmissivity and storativity for a two- dimensional, vertically confined aquifer, such as a long vertically fractured zone of high permeability within low permeable rocks or a long, high-permeability trench inside a low-permeability porous media. In addition, it can be used to analyze time-drawdown responses to pumping and injection in similar settings. The solution can also be used to approximate the groundwater flow for unconfined conditions if (1) the variation of transmissivity is negligible (groundwater table variation is small in comparison to the saturated thickness); and (2) the unsaturated flow is negligible. The errors associated with the use of the solution to unconfined conditions depend on the accuracies of the above two assumptions. The solution can also be used to assess the impacts of recharge from a seasonal river or irrigation canal on the groundwater system by assuming uniform, time- constant recharge along the river or canal. This paper presents the details for derivation of the analytical solution. The analytical solution is compared to numerical simulation results with example cases. Its accuracy is also assessed and
Hydrogen generation from aluminium corrosion in reactor containment spray solutions
International Nuclear Information System (INIS)
Frid, W.; Karlberg, G.; Sundvall, S.B.
1982-01-01
The aluminium corrosion experiments in reactor containment spray solutions, under the conditions expected to prevail during LOCA in BWR and PWR, were performed in order to investigate relationships between temperature, pH and hydrogen production rates. In order to simulate the conditions in a BWR containment realistic ratios between aluminium surface and water volume and between aluminium surface and oxygen volume were used. Three different aluminium alloys were exposed to spray solutions: AA 1050, AA 5052 and AA 6082. The corrosion rates were measured for BWR solutions (deaerated and aerated) with pH 5 and 9 at 50, 100 and 150 0 C. The pressure was constantly 0.8 MPa. The hydrogen production rate was measured by means of gas chromatography. In deionized BWR water the corrosion rates did not exceed about 0.05 mm/year in all cases, i.e. were practically independent of temperature and pH. Hydrogen concentrations were less than 0.1 vol.% in cooled dry gas. Corrosion rates and hydrogen production in PWR alkaline solution measured at pH 9.7 and 150 0 C were very high. AA 5052 alloy was the best material
Fast Reactor Systems and Innovative Fuels for Minor Actinides Homogeneous Recycling
International Nuclear Information System (INIS)
Calabrese, R.
2013-01-01
The capability of nuclear energy source to limit GHG emissions at a competitive cost is still a potential driver for its development in the near- and medium-term. The sustainability of nuclear energy is concerned by various issues such as the shortage of natural uranium resources, the management of steadily increasing inventories of spent nuclear fuel as well as competitiveness. Nuclear technology should be, for its societal acceptability, affordable, safe and featured by low proliferation risks. In this regard innovative fast reactors could improve the management of spent nuclear fuel inventories a reducing the burden on the geological repository. The development of MA-bearing oxide fuels is ongoing both on the definition of under-irradiation behaviour as well as the investigations of new fabrication routes and significant efforts in R&D are necessary. This paper confirms the expected performance of investigated FRs and the synergistic use of NFCSS and DESAE proved to be capable in modelling with reasonable accuracy an innovative fuel cycle strategy. The reduction of GHG emissions by means of a steep expansion of nuclear energy needs to be carefully investigated where a multi-criteria approach is of crucial importance
Numerical solutions of the aerosol general dynamic equation for nuclear reactor safety studies
International Nuclear Information System (INIS)
Park, J.W.
1988-01-01
Methods and approximations inherent in modeling of aerosol dynamics and evolution for nuclear reactor source term estimation have been investigated. Several aerosol evolution problems are considered to assess numerical methods of solving the aerosol dynamic equation. A new condensational growth model is constructed by generalizing Mason's formula to arbitrary particle sizes, and arbitrary accommodation of the condensing vapor and background gas at particle surface. Analytical solution is developed for the aerosol growth equation employing the new condensation model. The space-dependent aerosol dynamic equation is solved to assess implications of spatial homogenization of aerosol distributions. The results of our findings are as follows. The sectional method solving the aerosol dynamic equation is quite efficient in modeling of coagulation problems, but should be improved for simulation of strong condensation problems. The J-space transform method is accurate in modeling of condensation problems, but is very slow. For the situation considered, the new condensation model predicts slower aerosol growth than the corresponding isothermal model as well as Mason's model, the effect of partial accommodation is considerable on the particle evolution, and the effect of the energy accommodation coefficient is more pronounced than that of the mass accommodation coefficient. For the initial conditions considered, the space-dependent aerosol dynamics leads to results that are substantially different from those based on the spatially homogeneous aerosol dynamic equation
Research program and uses of the solution fueled reactor SILENE
International Nuclear Information System (INIS)
Barbry, F.; Ratel, R.
1985-09-01
Designed and operated by the Nuclear Protection and Safety Institute of the CEA, SILENE is an original small sized reactor fueled with an uranyl nitrate solution. The reactor is capable to operate in three modes: ''Pulse'' operation (high power levels up to 1000 Megawatts during several millisecond), ''Free evolution'' operation (simulation of criticality accident excursions), ''Steady state'' operation in a power range of 0.01 W to 1 kW. The core can be surrounded by appropriate shields (lead, polyethylene) to vary the leakage radiations and the gamma to neutron dose ratio. It's possible to insert in the central cavity of the annular core vessel some capsules, devices or samples to be submitted to very high radiations levels. The research activities are mainly devoted towards nuclear safety studies: the criticality accident studies, and the behavior of oxide fuels under transient conditions. Some examples of tests are presented. As to other applications of the SILENE facility, the main studies now in progress deal with: designing and calibration of Health physics intrumentation, neutron and gamma dosimetry, and, radiobiology. Once the characteristics of radiation field are qualified by calculations and experimental techniques, SILENE will be proposed as a reference source [fr
Slezak, Izabella H; Jasik-Slezak, Jolanta; Rogal, Mirosława; Slezak, Andrzej
2006-01-01
On the basis of model equation depending the membrane potential deltapsis, on mechanical pressure difference (deltaP), concentration polarization coefficient (zetas), concentration Rayleigh number (RC) and ratio concentration of solutions separated by membrane (Ch/Cl), the characteristics deltapsis = f(deltaP)zetas,RC,Ch/Cl for steady values of zetas, RC and Ch/Cl in single-membrane system were calculated. In this system neutral and isotropic polymeric membrane oriented in horizontal plane, the non-homogeneous binary electrolytic solutions of various concentrations were separated. Nonhomogeneity of solutions is results from creations of the concentration boundary layers on both sides of the membrane. Calculations were made for the case where on a one side of the membrane aqueous solution of NaCl at steady concentration 10(-3) mol x l(-1) (Cl) was placed and on the other aqueous solutions of NaCl at concentrations from 10(-3) mol x l(-1) to 2 x 10(-2) mol x l(-1) (Ch). Their densities were greater than NaCl solution's at 10(-3) mol x l(-1). It was shown that membrane potential depends on hydrodynamic state of a complex concentration boundary layer-membrane-concentration boundary layer, what is controlled by deltaP, Ch/Cl, RC and zetas.
General exact solution for homogeneous time-dependent self-gravitating perfect fluids
International Nuclear Information System (INIS)
Gaete, P.; Hojman, R.
1988-01-01
A procedure to obtain the general exact solution of Einstein equations for a self-gravitating spherically-symmetric static perfect fluid obeying an arbitrary equation of state, is applied to time-dependent Kantowsky-Sachs line elements (with spherical, planar and hyperbolic symmetry). As in the static case, the solution is generated by an arbitrary function of the independent variable and its first derivative. To illustrate the results, the whole family of (plane-symmetric) solutions with a ''gamma-law'' equation of state is explicity obtained in terms of simple known functions. It is also shown that, while in the static plane-symmtric line elements, every metric is in one to one correspondence with a ''partner-metric'' (both originated from the same generatrix function), in this case every generatrix function univocally determines one metric. (author) [pt
Chen, H.-C.
2016-01-01
The conversion and storage of solar energy into fuels provides a valuable solution for the future energy demand of our society. Making fuels via artificial photosynthesis, the so-called solar-to-fuel approach, is viewed as one of the most promising ways to produce clean and renewable energy.
Solutions to second order non-homogeneous multi-point BVPs using a fixed-point theorem
Directory of Open Access Journals (Sweden)
Yuji Liu
2008-07-01
Full Text Available In this article, we study five non-homogeneous multi-point boundary-value problems (BVPs of second order differential equations with the one-dimensional p-Laplacian. These problems have a common equation (in different function domains and different boundary conditions. We find conditions that guarantee the existence of at least three positive solutions. The results obtained generalize several known ones and are illustrated by examples. It is also shown that the approach for getting three positive solutions by using multi-fixed-point theorems can be extended to nonhomogeneous BVPs. The emphasis is on the nonhomogeneous boundary conditions and the nonlinear term involving first order derivative of the unknown. Some open problems are also proposed.
Energy Technology Data Exchange (ETDEWEB)
Hampden-Smith, M.; Kawola, J.S.; Martino, A.; Sault, A.G.; Yamanaka, S.A.
1999-01-05
The mission of this project was to use inverse micelle solutions to synthesize nanometer sized metal particles and test the particles as catalysts in the liquefaction of coal and other related reactions. The initial focus of the project was the synthesis of iron based materials in pseudo-homogeneous form. The frost three chapters discuss the synthesis, characterization, and catalyst testing in coal liquefaction and model coal liquefaction reactions of iron based pseudo-homogeneous materials. Later, we became interested in highly dispersed catalysts for coprocessing of coal and plastic waste. Bifunctional catalysts . to hydrogenate the coal and depolymerize the plastic waste are ideal. We began studying, based on our previously devised synthesis strategies, the synthesis of heterogeneous catalysts with a bifunctional nature. In chapter 4, we discuss the fundamental principles in heterogeneous catalysis synthesis with inverse micelle solutions. In chapter 5, we extend the synthesis of chapter 4 to practical systems and use the materials in catalyst testing. Finally in chapter 6, we return to iron and coal liquefaction now studied with the heterogeneous catalysts.
International Nuclear Information System (INIS)
Bouly, J.C.; Caizergues, R.; Deilgat, E.; Houelle, M.; Lecorche, P.
1967-01-01
This report groups together a series of experimental and theoretical studies on cylinders and plates of solution tried out at the Valduc Centre. a) Comparison of the theoretical and experimental results obtained on critical heights of solutions. b) Study of the effect of nitrogen, introduced in the form of the ion NO 3- , on the reactivity of fissile media. c) Study of the effect of 240 94 Pu on the reactivity of these media. d) Study of the influence of the dimensions of the inner cavity of annular cylinders, as well as of the influence of the moderator which may be introduced. Simple results were obtained which were easy to apply. An extrapolation to other geometries is made. (authors) [fr
Degradation of homogeneous polymer solutions in high shear turbulent pipe flow
Elbing, B. R.; Winkel, E. S.; Solomon, M. J.; Ceccio, S. L.
2009-12-01
This study quantifies degradation of polyethylene oxide (PEO) and polyacrylamide (PAM) polymer solutions in large diameter (2.72 cm) turbulent pipe flow at Reynolds numbers to 3 × 105 and shear rates greater than 105 1/s. The present results support a universal scaling law for polymer chain scission reported by Vanapalli et al. (2006) that predicts the maximum chain drag force to be proportional to Re 3/2, validating this scaling law at higher Reynolds numbers than prior studies. Use of this scaling gives estimated backbone bond strengths from PEO and PAM of 3.2 and 3.8 nN, respectively. Additionally, with the use of synthetic seawater as a solvent the onset of drag reduction occurred at higher shear rates relative to the pure water solvent solutions, but had little influence on the extent of degradation at higher shear rates. These results are significant for large diameter pipe flow applications that use polymers to reduce drag.
International Nuclear Information System (INIS)
Magat, Ph.
1997-04-01
Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P N method (SP N ) for the modeling of the core. The comparison between S N calculations and SP N calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P 3 , there is no effect; -) the P 1 development is adequate; and -) the P 0 development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP 1 , SP 3 , SP 5 and SP 7 ). These calculations show that the implementation of the SP N method is feasible but the extra-costs in computation times and memory are important. We recommend: SP 5 P 1 calculations for heterogeneous 2-dimension geometry and SP 3 P 1 calculations for the homogeneous 3-dimension geometry. (A.C.)
Classification of stable solutions for non-homogeneous higher-order elliptic PDEs
Directory of Open Access Journals (Sweden)
Abdellaziz Harrabi
2017-04-01
Full Text Available Abstract Under some assumptions on the nonlinearity f, we will study the nonexistence of nontrivial stable solutions or solutions which are stable outside a compact set of R n $\\mathbb {R}^{n}$ for the following semilinear higher-order problem: ( − Δ k u = f ( u in R n , $$\\begin{aligned} (-\\Delta^{k} u= f(u \\quad \\mbox{in }\\mathbb {R}^{n}, \\end{aligned}$$ with k = 1 , 2 , 3 , 4 $k=1,2,3,4$ . The main methods used are the integral estimates and the Pohozaev identity. Many classes of nonlinearity will be considered; even the sign-changing nonlinearity, which has an adequate subcritical growth at zero as for example f ( u = − m u + λ | u | θ − 1 u − μ | u | p − 1 u $f(u= -m u +\\lambda|u|^{\\theta-1}u-\\mu |u|^{p-1}u$ , where m ≥ 0 $m\\geq0$ , λ > 0 $\\lambda>0$ , μ > 0 $\\mu>0$ , p , θ > 1 $p, \\theta>1$ . More precisely, we shall revise the nonexistence theorem of Berestycki and Lions (Arch. Ration. Mech. Anal. 82:313-345, 1983 in the class of smooth finite Morse index solutions as the well known work of Bahri and Lions (Commun. Pure Appl. Math. 45:1205-1215, 1992. Also, the case when f ( u u $f(uu$ is a nonnegative function will be studied under a large subcritical growth assumption at zero, for example f ( u = | u | θ − 1 u ( 1 + | u | q $f(u=|u|^{\\theta-1}u(1 + |u|^{q}$ or f ( u = | u | θ − 1 u e | u | q $f(u= |u|^{\\theta-1}u e^{|u|^{q}}$ , θ > 1 $\\theta>1$ and q > 0 $q>0$ . Extensions to solutions which are merely stable are discussed in the case of supercritical growth with k = 1 $k=1$ .
Solutions of Boltzmann`s Equation for Mono-energetic Neutrons in an Infinite Homogeneous Medium
Wigner, E. P.
1943-11-30
Boltzman's equation is solved for the case of monoenergetic neutrons created by a plane or point source in an infinite medium which has spherically symmetric scattering. The customary solution of the diffusion equation appears to be multiplied by a constant factor which is smaller than 1. In addition to this term the total neutron density contains another term which is important in the neighborhood of the source. It varies as 1/r{sup 2} in the neighborhood of a point source. (auth)
International Nuclear Information System (INIS)
Bezerra, Jair de Lima; Lira, Carlos Alberto Brayner de Oliveira; Barroso, Antonio Carlos de Oliveira; Lima, Fernando Roberto de Andrade; Bezerra da Silva, Mário Augusto
2013-01-01
Highlights: ► Experimental bench with test section made of transparent acrylic, simulating the pressurizer reactor IRIS. ► Workbench used to study the process of homogenization of boron in the pressurizer IRIS nuclear reactor. ► Results were obtained through videos and digital photos of the test section. - Abstract: The reactivity control of a nuclear reactor to pressurized water is made by means of controlling bars or by boron dilution in the water from the coolant of a primary circuit. The control with boron dilution has great importance, despite inserting small variations in the reactivity in the reactor, as it does not significantly affect the distribution of the neutron flux. A simplified experimental bench with a test section manufactured in transparent acrylic, was built in reduced scale as to be used in a boron homogenizing process, simulating an IRIS reactor pressurizer (International Reactor Innovative and Secure). The bench was assembled in the Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE), an entity linked to the Comissão Nacional de Energia Nuclear (CNEN), Recife – PE
International Nuclear Information System (INIS)
Bezerra, Jair de Lima; Lira, Carlos Alberto Brayner de Oliveira; Barroso, Antonio Carlos de Oliveira; Lima, Fernando Roberto de Andrade; Silva, Mário Augusto Bezerra da
2013-01-01
Highlights: • Experimental bench with test section made of transparent acrylic, simulating the pressurizer reactor IRIS. • Workbench used to study the process of homogenization of boron in the pressurizer IRIS nuclear reactor. • Results were obtained through videos and digital photos of the test section. - Abstract: The reactivity control of a nuclear reactor to pressurized water is made by means of controlling bars or by boron dilution in the water from the coolant of a primary circuit. The control with boron dilution has great importance, despite inserting small variations in the reactivity in the reactor, as it does not significantly affect the distribution of the neutron flux. A simplified experimental bench with a test section manufactured in transparent acrylic, was built in reduced scale as to be used in a boron homogenizing process, simulating an IRIS reactor pressurizer (International Reactor Innovative and Secure). The bench was assembled in the Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE), an entity linked to the Comissão Nacional de Energia Nuclear (CNEN), Recife–PE
Time-Homogeneous Parabolic Wick-Anderson Model in One Space Dimension: Regularity of Solution
Kim, Hyun-Jung; Lototsky, Sergey V
2017-01-01
Even though the heat equation with random potential is a well-studied object, the particular case of time-independent Gaussian white noise in one space dimension has yet to receive the attention it deserves. The paper investigates the stochastic heat equation with space-only Gaussian white noise on a bounded interval. The main result is that the space-time regularity of the solution is the same for additive noise and for multiplicative noise in the Wick-It\\^o-Skorokhod interpretation.
International Nuclear Information System (INIS)
Patra, A.; Saha Ray, S.
2014-01-01
Highlights: • A stationary transport equation has been solved using the technique of Haar wavelet Collocation Method. • This paper intends to provide the great utility of Haar wavelets to nuclear science problem. • In the present paper, two-dimensional Haar wavelets are applied. • The proposed method is mathematically very simple, easy and fast. - Abstract: This paper emphasizes on finding the solution for a stationary transport equation using the technique of Haar wavelet Collocation Method (HWCM). Haar wavelet Collocation Method is efficient and powerful in solving wide class of linear and nonlinear differential equations. Recently Haar wavelet transform has gained the reputation of being a very effective tool for many practical applications. This paper intends to provide the great utility of Haar wavelets to nuclear science problem. In the present paper, two-dimensional Haar wavelets are applied for solution of the stationary Neutron Transport Equation in homogeneous isotropic medium. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency of the method, one test problem is discussed. It can be observed from the computational simulation that the numerical approximate solution is much closer to the exact solution
International Nuclear Information System (INIS)
Foroutan, A.
1992-05-01
The essential mathematical challenge in transport theory is based on the nonlinearity of the integro-differential equations governing classical thermodynamic systems on molecular kinetic level. It is the aim of this thesis to gain exact analytical solutions to the model Boltzmann equation suggested by Tjon and Wu. Such solutions afford a deeper insight into the dynamics of rarefied gases. Tjon and Wu have provided a stochastic model of a Boltzmann equation. Its transition probability depends only on the relative speed of the colliding particles. This assumption leads in the case of two translational degrees of freedom to an integro-differential equation of convolution type. According to this convolution structure the integro-differential equation is Laplace transformed. The result is a nonlinear partial differential equation. The investigation of the symmetries of this differential equation by means of Lie groups of transformation enables us to transform the originally nonlinear partial differential equation into ordinary differential equation into ordinary differential equations of Bernoulli type. (author)
Krupež, Jelena; Kovačević, Vesna V.; Jović, Milica; Roglić, Goran M.; Natić, Maja M.; Kuraica, Milorad M.; Obradović, Bratislav M.; Dojčinović, Biljana P.
2018-05-01
Nicotine degradation efficiency in water solutions was studied using a water falling film dielectric barrier discharge (DBD) reactor. Two different treatments were applied: direct treatment, the recirculation of the solution through a DBD reactor, and indirect treatment, the bubbling of the gas from the DBD through the porous filter into the solution. In a separate experiment, samples spiked with nicotine in double distilled water (ddH2O) and tap water were studied and compared after both treatments. Furthermore, the effects of the homogeneous catalysts, namely, Fe2+ and H2O2, were tested in the direct treatment. Nicotine degradation efficiency was determined using high-performance liquid chromatography. A degradation efficiency of 90% was achieved after the direct treatment catalyzed with Fe2+. In order to analyze the biodegradability, mineralization level, and toxicity of the obtained solutions, after all degradation procedures the values of the following parameters were determined: total organic carbon, chemical oxygen demand, biochemical oxygen demand, and the Artemia salina toxicity test. The results showed that an increase in biodegradability was obtained, after all treatments. A partial nicotine mineralization was achieved and the mortality of the A. salina organism decreased in the treated samples, all of which indicating the effective removal of nicotine and the creation of less toxic solutions. Nicotine degradation products were identified using ultrahigh-performance liquid chromatography coupled with a linear ion trap Orbitrap hybrid mass spectrometer and a simple mechanism for oxidative degradation of nicotine in non-thermal plasma systems is proposed.
Directory of Open Access Journals (Sweden)
T. Deshler
2017-06-01
Full Text Available Ozone plays a significant role in the chemical and radiative state of the atmosphere. For this reason there are many instruments used to measure ozone from the ground, from space, and from balloons. Balloon-borne electrochemical cell ozonesondes provide some of the best measurements of the ozone profile up to the mid-stratosphere, providing high vertical resolution, high precision, and a wide geographic distribution. From the mid-1990s to the late 2000s the consistency of long-term records from balloon-borne ozonesondes has been compromised by differences in manufacturers, Science Pump (SP and ENSCI (EN, and differences in recommended sensor solution concentrations, 1.0 % potassium iodide (KI and the one-half dilution: 0.5 %. To investigate these differences, a number of organizations have independently undertaken comparisons of the various ozonesonde types and solution concentrations, resulting in 197 ozonesonde comparison profiles. The goal of this study is to derive transfer functions to allow measurements outside of standard recommendations, for sensor composition and ozonesonde type, to be converted to a standard measurement and thus homogenize the data to the expected accuracy of 5 % (10 % in the stratosphere (troposphere. Subsets of these data have been analyzed previously and intermediate transfer functions derived. Here all the comparison data are analyzed to compare (1 differences in sensor solution composition for a single ozonesonde type, (2 differences in ozonesonde type for a single sensor solution composition, and (3 the World Meteorological Organization's (WMO and manufacturers' recommendations of 1.0 % KI solution for Science Pump and 0.5 % KI for ENSCI. From the recommendations it is clear that ENSCI ozonesondes and 1.0 % KI solution result in higher amounts of ozone sensed. The results indicate that differences in solution composition and in ozonesonde type display little pressure dependence at pressures
Radhakrishnan, Krishnan
1994-01-01
LSENS, the Lewis General Chemical Kinetics and Sensitivity Analysis Code, has been developed for solving complex, homogeneous, gas-phase chemical kinetics problems and contains sensitivity analysis for a variety of problems, including nonisothermal situations. This report is part 1 of a series of three reference publications that describe LENS, provide a detailed guide to its usage, and present many example problems. Part 1 derives the governing equations and describes the numerical solution procedures for the types of problems that can be solved. The accuracy and efficiency of LSENS are examined by means of various test problems, and comparisons with other methods and codes are presented. LSENS is a flexible, convenient, accurate, and efficient solver for chemical reaction problems such as static system; steady, one-dimensional, inviscid flow; reaction behind incident shock wave, including boundary layer correction; and perfectly stirred (highly backmixed) reactor. In addition, the chemical equilibrium state can be computed for the following assigned states: temperature and pressure, enthalpy and pressure, temperature and volume, and internal energy and volume. For static problems the code computes the sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of the dependent variables and/or the three rate coefficient parameters of the chemical reactions.
International Nuclear Information System (INIS)
Moline, G.R.
1998-03-01
The Homogeneous Reactor Experiment (HRE) Pond is the site of a former impoundment for radioactive wastes that has since been drained, filled with soil, and covered with an asphalt cap. The site is bordered to the east and south by a tributary that empties into Melton Branch Creek and that contains significant concentrations of radioactive contaminants, primarily 90 Sr. Because of the proximity of the tributary to the HRE disposal site and the probable flow of groundwater from the site to the tributary, it is hypothesized that the HRE Pond is a source of contamination to he creek. As a means for temporary containment of contaminants within the impoundment, a cryogenic barrier technology demonstration was initiated in FY96 with a background hydrologic investigation that continued through FY97. Cryogenic equipment installation was completed in FY97, and freezing was initiated in September of 1997. This report documents the results of a hydrologic and geologic investigation of the HRE Pond/cryogenic barrier site. The purpose of this investigation is to evaluate the hydrologic conditions within and around the impoundment in order to meet the following objectives: (1) to provide a pre-barrier subsurface hydrologic baseline for post-barrier performance assessment; (2) to confirm that the impoundment is hydraulically connected to the surrounding sediments; and (3) to determine the likely contaminant exit pathways from the impoundment. The methods of investigation included water level and temperature monitoring in a network of wells and standpipes in and surrounding the impoundment, a helium tracer test conducted under ambient flow conditions, and geologic logging during the drilling of boreholes for installation of cryogenic probes and temperature monitoring wells
Fast reactors as a solution for future small-scale nuclear energy
International Nuclear Information System (INIS)
Kudryavtseva, A.; Danilenko, K.; Dorofeev, K.
2013-01-01
Small nuclear power plants can provide a future platform for decentralized energy supply providing better levels of accessibility, safety and environmental friendliness. The optimal solution for SMR deployment is fast reactors with inherent safety. To compete alternative solutions SMRs must exhibit some evident advantages in: safety, technology, and economic. Small modular reactors with lead-bismuth coolant (SVBR-100) under development in Russia can be a prospective solution for future small and decentralized energy
Hage, Ilige S; Hamade, Ramsey F
2017-09-01
solution are corroborated experimentally using microhardness indentation measurements taken at the same points that the digital images were taken along a radial distance emanating from the interior (endosteum) surface toward the bone's exterior (periosteum) surface. Good agreement was found between numerically calculated and indentation measured stiffness of Intracortical lamellae. Both indentation measurements and numerical solutions of matrix stiffness showed increasing linear trend of compressive longitudinal modulus (E11) values vs. radial position for both interior and exterior regions. In the interior (exterior) region of cortical bone, stiffness modulus values were found to range from 18.5 to 23.4 GPa (23 to 26.0 GPa) with the aggregate stiffness of the cortical lamella in the exterior region being 12% stiffer than that in the interior region. In order to further validate these findings, experimental and FEM simulation of a mid-diaphysis bone ring under compression is employed. The FEM numerical deflections employed nine concentric regions across the thickness with graded stiffness values based on the digital segmentation and homogenization scheme. Bone ring deflections are found to agree well with measured deformations of the compression bone ring.
International Nuclear Information System (INIS)
Zhotabaev, Zh.R.; Solov'ev, Yu.A.
2001-01-01
The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed
An alternative solution for heavy liquid metal cooled reactors fuel assemblies
Energy Technology Data Exchange (ETDEWEB)
Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)
2014-10-15
Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.
Analytic solutions of the multigroup space-time reactor kinetics equations
International Nuclear Information System (INIS)
Lee, C.E.; Rottler, S.
1986-01-01
The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)
International Nuclear Information System (INIS)
Ujihira, Yusuke; Ohyabu, Matashige; Murakami, Tetsuro; Horie, Tsuyoshi.
1978-01-01
Chemical states of iron(III) compounds, precipitated homogeneously by heating the iron(III) salt solution at 363 K in the presence of urea, was studied by means of Moessbauer spectrometry and X-ray diffractometry. The pH-time relation of urea hydrolysis revealed that the precipitation process from homogeneous solution is identical to the hydrolysis of iron(III) ion at pH around 2 under the homogeneous supply of OH - ion, which is generated by hydrolysis of urea. Accordingly, iron(III) oxide hydroxide or similar compounds to the hydrolysis products of iron(III) ion was precipitated by the precipitation from homogeneous solution methods. Akaganeite (β-FeOOH) was crystallized from 0.1 M iron(III) chloride solution. Goethite(α-FeOOH) and hematite(α-Fe 2 O 3 ) was precipitated from 0.1 M iron(III) nitrate solution, vigorous liberation of OH - ion favoring the crystallization of hematite. The addition of chloride ion to the solution resulted in the formation of akaganeite. Basic salt of iron sulfate[NH 4 Fe 3 (OH) 6 (SO 4 ) 2 ] and goethite were formed from 0.1 M iron(III) sulfate solution, the former being obtained in the more moderate condition of the urea hydrolysis ( 363 K). (author)
Homogenization of neutronic diffusion models
International Nuclear Information System (INIS)
Capdebosq, Y.
1999-09-01
In order to study and simulate nuclear reactor cores, one needs to access the neutron distribution in the core. In practice, the description of this density of neutrons is given by a system of diffusion equations, coupled by non differential exchange terms. The strong heterogeneity of the medium constitutes a major obstacle to the numerical computation of this models at reasonable cost. Homogenization appears as compulsory. Heuristic methods have been developed since the origin by nuclear physicists, under a periodicity assumption on the coefficients. They consist in doing a fine computation one a single periodicity cell, to solve the system on the whole domain with homogeneous coefficients, and to reconstruct the neutron density by multiplying the solutions of the two computations. The objectives of this work are to provide mathematically rigorous basis to this factorization method, to obtain the exact formulas of the homogenized coefficients, and to start on geometries where two periodical medium are placed side by side. The first result of this thesis concerns eigenvalue problem models which are used to characterize the state of criticality of the reactor, under a symmetry assumption on the coefficients. The convergence of the homogenization process is proved, and formulas of the homogenized coefficients are given. We then show that without symmetry assumptions, a drift phenomenon appears. It is characterized by the mean of a real Bloch wave method, which gives the homogenized limit in the general case. These results for the critical problem are then adapted to the evolution model. Finally, the homogenization of the critical problem in the case of two side by side periodic medium is studied on a one dimensional on equation model. (authors)
Numerical solutions to critical problem of reflected cylindrical reactor
International Nuclear Information System (INIS)
Horie, Junnosuke
1977-01-01
The multi-region critical problem can be transformed into an eigenvalue problem in the classical sense by using the method of Kuscer and Corngold and of Wing. This transformation is applied to derive a variational formulation for a reflected reactor. An approximate critical value of the multiplying factor is determined by maximizing the Rayleigh quotient for radially and totally reflected cylindrical reactors. It is shown that this approximate critical value is an upper bound of the true critical value. From the facts that the operator is self-adjoint and the eigenfunction is positive, an expression is derived for the upper and lower bounds of the true eigenvalue, by making use of the approximate distribution. The difference of the upper and lower bounds is an uncertainty of the presumption of the true critical value. It is found that we can compute the bounds to any required precision. The narrow bounds are calculated for two radially and one totally reflected cylindrical reactors. (auth.)
Mines vs reactors: comparison of radiation problems and solutions
International Nuclear Information System (INIS)
Bush, W.R.
1981-01-01
Radiation protection in uranium mines and nuclear reactors is compared, with the objective of determining if any radiation protection practices which have worked well in one area can be adapted to the other area, especially from reactors to mines since radiation protection is more highly developed in reactors. Several areas are identified where improvements can be made in radiation protection in mines, the most important being improvement in ventilation at the workplace, greater attention to special monitoring during upset conditions, and a need for a substantial increase in radiation protection staff. It is concluded that, although considerable room exists for improvement, the general approach to radiation protection presently being followed seems appropriate for the circumstances encountered in mining
Reactor cavity streaming: the problem and engineered solutions
International Nuclear Information System (INIS)
Iotti, R.C.; Yang, T.L.; Rogers, W.H.
1979-01-01
Experience at operating pressurized water reactors has revealed that air gaps between the reactor vessel and the biological shield wall can provide paths for radiation streaming, which may prohibitively limit the accessibility required to areas in the containment during power operation, increase personnel exposure during shutdown, and cause radiation damage to equipment and cables located above the vessel. Several concepts of shield are discussed together with their predicted effectiveness. The analytical methods employed to determine the streaming magnitude and the shield effectiveness are also discussed and their accuracy is measured by comparison with actual measurement at an operating plant
Stochastic solution of population balance equations for reactor networks
International Nuclear Information System (INIS)
Menz, William J.; Akroyd, Jethro; Kraft, Markus
2014-01-01
This work presents a sequential modular approach to solve a generic network of reactors with a population balance model using a stochastic numerical method. Full-coupling to the gas-phase is achieved through operator-splitting. The convergence of the stochastic particle algorithm in test networks is evaluated as a function of network size, recycle fraction and numerical parameters. These test cases are used to identify methods through which systematic and statistical error may be reduced, including by use of stochastic weighted algorithms. The optimal algorithm was subsequently used to solve a one-dimensional example of silicon nanoparticle synthesis using a multivariate particle model. This example demonstrated the power of stochastic methods in resolving particle structure by investigating the transient and spatial evolution of primary polydispersity, degree of sintering and TEM-style images. Highlights: •An algorithm is presented to solve reactor networks with a population balance model. •A stochastic method is used to solve the population balance equations. •The convergence and efficiency of the reported algorithms are evaluated. •The algorithm is applied to simulate silicon nanoparticle synthesis in a 1D reactor. •Particle structure is reported as a function of reactor length and time
Saien, Javad; Fallah Vahed Bazkiaei, Marzieh
2018-07-01
Aqueous solutions of p-nitrophenol (PNP) were treated with UV-activated potassium periodate (UV/KPI) in an efficient photo-reactor. Either periodate or UV alone had little effect; however, their combination led to a significant degradation and mineralization. The response surface methodology was employed for design of experiments and optimization. The optimum conditions for treatment of 30 mg/L of the substrate were determined as [KPI] = 386.3 mg/L, pH = 6.2 and T = 34.6°C, under which 79.5% degradation was achieved after 60 min. Use of 25 and 40 kHz ultrasound waves caused the degradation to enhance to 88.3% and 92.3%, respectively. The intermediates were identified by gas chromatography-mass spectroscopy analysis, leading to propose the reaction pathway. The presence of water conventional bicarbonate, chloride, sulfate and nitrate anions caused unfavorable effects in efficiency. Meanwhile, the kinetic study showed that PNP degradation follows a pseudo-first-order reaction and the activation energy was determined. The irradiation energy consumption required for one order of magnitude degradation was estimated as 11.18 kWh/m 3 . Accordingly, comparison with the previously reported processes showed the superiority of PNP treatment with the employed process.
Numerical solution of the kinetic equation in reactor shielding
International Nuclear Information System (INIS)
Germogenova, T.A.
1975-01-01
A review is made of methods of solving marginal problems of multi-group systems of equations of neutron and γ radiation transfer. The first stage of the solution - the quantification of the basic task, is determined by the qualitative behaviour of the solution - is the nature of its performance and asymptotics. In the second stage - solution of the approximating system, various modifications of the iterative method are as a rule used. A description is given of the features of the major Soviet complexes of programmes (ROZ and RADUGA) for the solution of multi-group systems of transfer equations and some methodological research findings are presented. (author)
Hybrid diffusion–transport spatial homogenization method
International Nuclear Information System (INIS)
Kooreman, Gabriel; Rahnema, Farzad
2014-01-01
Highlights: • A new hybrid diffusion–transport homogenization method. • An extension of the consistent spatial homogenization (CSH) transport method. • Auxiliary cross section makes homogenized diffusion consistent with heterogeneous diffusion. • An on-the-fly re-homogenization in transport. • The method is faster than fine-mesh transport by 6–8 times. - Abstract: A new hybrid diffusion–transport homogenization method has been developed by extending the consistent spatial homogenization (CSH) transport method to include diffusion theory. As in the CSH method, an “auxiliary cross section” term is introduced into the source term, making the resulting homogenized diffusion equation consistent with its heterogeneous counterpart. The method then utilizes an on-the-fly re-homogenization in transport theory at the assembly level in order to correct for core environment effects on the homogenized cross sections and the auxiliary cross section. The method has been derived in general geometry and tested in a 1-D boiling water reactor (BWR) core benchmark problem for both controlled and uncontrolled configurations. The method has been shown to converge to the reference solution with less than 1.7% average flux error in less than one third the computational time as the CSH method – 6 to 8 times faster than fine-mesh transport
International Nuclear Information System (INIS)
Tzanos, C.P.
1980-01-01
The differences in fissile inventory and breeding ratio, with respect to the differences in fertile inventory and neutron spectrum, between equivalent heterogeneous and homogeneous configurations were analyzed. To quantify the effect of spectral changes on reaction rate ratios, a calculational scheme based on properly prepared one-group cross-section sets was used
International Nuclear Information System (INIS)
Zaza, Chady
2015-01-01
The numerical simulation of steam generators of pressurized water reactors is a complex problem, involving different flow regimes and a wide range of length and time scales. An accidental scenario may be associated with very fast variations of the flow with an important Mach number. In contrast in the nominal regime the flow may be stationary, at low Mach number. Moreover whatever the regime under consideration, the array of U-tubes is modelled by a porous medium in order to avoid taking into account the complex geometry of the steam generator, which entails the issue of the coupling conditions at the interface with the free-fluid. We propose a new pressure-correction scheme for cell-centered finite volumes for solving the compressible Navier-Stokes and Euler equations at all Mach number. The existence of a discrete solution, the consistency of the scheme in the Lax sense and the positivity of the internal energy were proved. Then the scheme was extended to the homogeneous two-phase flow models of the GENEPI code developed at CEA. Lastly a multigrid-AMR algorithm was adapted for using our pressure-correction scheme on adaptive grids. Regarding the second issue addressed in this work, the numerical simulation of a fluid flow over a porous bed involves very different length scales. Macroscopic interface models - such as Ochoa-Tapia-Whitaker or Beavers-Joseph law for a viscous flow - represent the transition region between the free-fluid and the porous region by an interface of discontinuity associated with specific transmission conditions. An extension to the Beavers-Joseph law was proposed for the convective regime. By introducing a jump in the kinetic energy at the interface, we recover an interface condition close to the Beavers-Joseph law but with a non-linear slip coefficient, which depends on the free-fluid velocity at the interface and on the Darcy velocity. The validity of this new transmission condition was assessed with direct numerical simulations at
An iterative homogenization technique that preserves assembly core exchanges
International Nuclear Information System (INIS)
Mondot, Ph.; Sanchez, R.
2003-01-01
A new interactive homogenization procedure for reactor core calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly type is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macro-group assembly exchanges in the core. This homogenization avoids the periodic lattice-leakage model approximation and gives detailed assembly transport fluxes without need of an approximated flux reconstruction. Preliminary results are given for a one-dimensional core model. (authors)
International Nuclear Information System (INIS)
Tullett, J.D.
1990-01-01
P Benoist has developed a method for calculating cross-sections for Fast Reactor control rods and their followers described by a single homogenised region (the Equivalent Parameter Method). When used in a diffusion theory calculation, these equivalent cross-sections should give the same rod worth as one would obtain from a transport theory calculation with a heterogeneous description of the control rod and the follower. In this report, Benoist's theory is described, and a comprehensive set of tests is presented. These tests show that the method gives very good results over a range of geometries and control rod positions for a model fast reactor core. (author)
Energy Technology Data Exchange (ETDEWEB)
Stefanovic, D B [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1970-07-01
The objective of this work is to describe the new analytical solution of the neutron slowing down equation for infinite monoatomic media with arbitrary energy dependence of cross section. The solution is obtained by introducing Green slowing down functions instead of starting from slowing down equations directly. The previously used methods for calculation of fission neutron spectra in the reactor cell were numerical. The proposed analytical method was used for calculating the space-energy distribution of fast neutrons and number of neutron reactions in a thermal reactor cell. The role of analytical method in solving the neutron slowing down in reactor physics is to enable understating of the slowing down process and neutron transport. The obtained results could be used as standards for testing the accuracy od approximative and practical methods.
Hydrogen evolution from aluminium in reactor containment spray solutions
International Nuclear Information System (INIS)
Karlberg, G.; Sundvall, S.-B.
1982-01-01
Three different aluminium alloys were exposed to conditions similar to BWR and PWR containment spray waters at 50, 100 and 150 0 C. BWR deionized water gives corrosion rates of at most 0.05 mm/year and hydrogen concentrations less than 0.1-1%. On the contrary PWR alkaline solutions give very high corrosion rates and hydrogen contents. (Auth.)
Energy Technology Data Exchange (ETDEWEB)
Delgado H, C. E.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: ce_delgado89@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)
2015-10-15
Full text: One of the energy alternatives to fossil fuels which do not produce greenhouse gases is the nuclear energy. One of the drawbacks of this alternative is the generation of radioactive wastes of long half-life and its relation to the generation of nuclear materials to produce weapons of mass destruction. An option to these drawbacks of nuclear energy is to use Thorium as part of the nuclear fuel which it becomes in U{sup 233} when capturing neutrons, that is a fissile material. In this paper Monte Carlo methods were used to design a homogeneous subcritical reactor based on thorium. As neutron reflector graphite was used. The reactor core is homogeneous and is formed of 70% light water as moderator, 12% of enriched uranium UO{sub 2}(NO{sub 3}){sub 4} and 18% of thorium Th(NO{sub 3}){sub 4} as fuel. To start the nuclear fission chain reaction an isotopic source of californium 252 was used with an intensity of 4.6 x 10{sup 7} s{sup -1}. In the design the value of the effective multiplication factor, whose value turned out k{sub eff} <1 was calculated. Also, the neutron spectra at different distances from the source and the total fluence were calculated, as well as the values of the ambient dose equivalent in the periphery of the reactor. (Author)
International Nuclear Information System (INIS)
Demidenko, Eugene
2011-01-01
An analytic solution of the potential distribution on a 2D homogeneous disk for electrical impedance tomography under the complete electrode model is expressed via an infinite system of linear equations. For the shunt electrode model with two electrodes, our solution coincides with the previously derived solution expressed via elliptic integral (Pidcock et al 1995 Physiol. Meas. 16 77–90). The Dirichlet-to-Neumann map is derived for statistical estimation via nonlinear least squares. The solution is validated in phantom experiments and applied for breast contact impedance estimation in vivo. Statistical hypothesis testing is used to test whether the contact impedances are the same across electrodes or all equal zero. Our solution can be especially useful for a rapid real-time test for bad surface contact in clinical setting
Approximate solution for the reactor neutron probability distribution
International Nuclear Information System (INIS)
Ruby, L.; McSwine, T.L.
1985-01-01
Several authors have studied the Kolmogorov equation for a fission-driven chain-reacting system, written in terms of the generating function G(x,y,z,t) where x, y, and z are dummy variables referring to the neutron, delayed neutron precursor, and detector-count populations, n, m, and c, respectively. Pal and Zolotukhin and Mogil'ner have shown that if delayed neutrons are neglected, the solution is approximately negative binomial for the neutron population. Wang and Ruby have shown that if the detector effect is neglected, the solution, including the effect of delayed neutrons, is approximately negative binomial. All of the authors assumed prompt-neutron emission not exceeding two neutrons per fission. An approximate method of separating the detector effect from the statistics of the neutron and precursor populations has been proposed by Ruby. In this weak-coupling limit, it is assumed that G(x,y,z,t) = H(x,y)I(z,t). Substitution of this assumption into the Kolmogorov equation separates the latter into two equations, one for H(x,y) and the other for I(z,t). Solution of the latter then gives a generating function, which indicates that in the weak-coupling limit, the detector counts are Poisson distributed. Ruby also showed that if the detector effect is neglected in the equation for H(x,y), i.e., the detector efficiency is set to zero, then the resulting equation is identical with that considered by Wang and Ruby. The authors present here an approximate solution for H(x,y) that does not set the detector efficiency to zero
Czech Academy of Sciences Publication Activity Database
Šubrt, Jan; Štengl, Václav; Bakardjieva, Snejana; Szatmáry, Lórant
2006-01-01
Roč. 169, č. 1 (2006), s. 33-40 ISSN 0032-5910 R&D Projects: GA MŠk LC523; GA ČR GA104/04/0467 Institutional research plan: CEZ:AV0Z40320502 Keywords : nanoparticles * homogenous hydrolysis * oxides Subject RIV: CA - Inorganic Chemistry Impact factor: 1.232, year: 2006
International Nuclear Information System (INIS)
Barbry, F.
1984-01-01
After defining the content and the objectives of criticality accident studies, the SILENE reactor, a means of studying fuel solution criticality accidents, is presented. Information obtained from the CRAC and SILENE experimental programs are then presented; they concern power excursion phenomenology, radiological consequences, and finally guide-lines for current and future programs
AQUEOUS HOMOGENEOUS REACTORTECHNICAL PANEL REPORT
Energy Technology Data Exchange (ETDEWEB)
Diamond, D.J.; Bajorek, S.; Bakel, A.; Flanagan, G.; Mubayi, V.; Skarda, R.; Staudenmeier, J.; Taiwo, T.; Tonoike, K.; Tripp, C.; Wei, T.; Yarsky, P.
2010-12-03
Considerable interest has been expressed for developing a stable U.S. production capacity for medical isotopes and particularly for molybdenum- 99 (99Mo). This is motivated by recent re-ductions in production and supply worldwide. Consistent with U.S. nonproliferation objectives, any new production capability should not use highly enriched uranium fuel or targets. Conse-quently, Aqueous Homogeneous Reactors (AHRs) are under consideration for potential 99Mo production using low-enriched uranium. Although the Nuclear Regulatory Commission (NRC) has guidance to facilitate the licensing process for non-power reactors, that guidance is focused on reactors with fixed, solid fuel and hence, not applicable to an AHR. A panel was convened to study the technical issues associated with normal operation and potential transients and accidents of an AHR that might be designed for isotope production. The panel has produced the requisite AHR licensing guidance for three chapters that exist now for non-power reactor licensing: Reac-tor Description, Reactor Coolant Systems, and Accident Analysis. The guidance is in two parts for each chapter: 1) standard format and content a licensee would use and 2) the standard review plan the NRC staff would use. This guidance takes into account the unique features of an AHR such as the fuel being in solution; the fission product barriers being the vessel and attached systems; the production and release of radiolytic and fission product gases and their impact on operations and their control by a gas management system; and the movement of fuel into and out of the reactor vessel.
Iterative solution to the optimal poison management problem in pressurized water reactors
International Nuclear Information System (INIS)
Colletti, J.P.; Levine, S.H.; Lewis, J.B.
1983-01-01
A new method for solving the optimal poison management problem for a multiregion pressurized water reactor has been developed. The optimization objective is to maximize the end-of-cycle core excess reactivity for any given beginning-of-cycle fuel loading. The problem is treated as an optimal control problem with the region burnup and control absorber concentrations acting as the state and control variables, respectively. Constraints are placed on the power peaking, soluble boron concentration, and control absorber concentrations. The solution method consists of successive relinearizations of the system equations resulting in a sequence of nonlinear programming problems whose solutions converge to the desired optimal control solution. Application of the method to several test problems based on a simplified three-region reactor suggests a bang-bang optimal control strategy with the peak power location switching between the inner and outer regions of the core and the critical soluble boron concentration as low as possible throughout the cycle
International Nuclear Information System (INIS)
Primm, R.T. III; Mincey, J.F.
1982-01-01
The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations
Ribeiro, F B
1999-01-01
Solutions of the diffusion equation in cylindrical coordinates are presented for a radionuclide produced by the decay of a not diffusing parent isotope with arbitrary activity distribution. General initial and Dirichlet boundary conditions are considered and the diffusion equation is solved for a finite cylinder. Solutions corresponding to two particular boundary conditions that can be imposed in laboratory diffusion coefficient measurements are presented. An analysis of the speed of convergence and of the series truncation error is done for these particular solutions. An example of the escape to production ratio derived from one of the solutions is also presented.
International Nuclear Information System (INIS)
Ribeiro, Fernando Brenha
1999-01-01
Solutions of the diffusion equation in cylindrical coordinates are presented for a radionuclide produced by the decay of a not diffusing parent isotope with arbitrary activity distribution. General initial and Dirichlet boundary conditions are considered and the diffusion equation is solved for a finite cylinder. Solutions corresponding to two particular boundary conditions that can be imposed in laboratory diffusion coefficient measurements are presented. An analysis of the speed of convergence and of the series truncation error is done for these particular solutions. An example of the escape to production ratio derived from one of the solutions is also presented
Finding Solutions to Different Problems Simultaneously in a Multi-molecule Simulated Reactor
Directory of Open Access Journals (Sweden)
Jaderick P. Pabico
2014-12-01
Full Text Available – In recent years, the chemical metaphor has emerged as a computational paradigm based on the observation of different researchers that the chemical systems of living organisms possess inherent computational properties. In this metaphor, artificial molecules are considered as data or solutions, while the interactions among molecules are defined by an algorithm. In recent studies, the chemical metaphor was used as a distributed stochastic algorithm that simulates an abstract reactor to solve the traveling salesperson problem (TSP. Here, the artificial molecules represent Hamiltonian cycles, while the reactor is governed by reactions that can re-order Hamiltonian cycles. In this paper, a multi-molecule reactor (MMR-n that simulates chemical catalysis is introduced. The MMR-n solves in parallel three NP-hard computational problems namely, the optimization of the genetic parameters of a plant growth simulation model, the solution to large instances of symmetric and asymmetric TSP, and the static aircraft landing scheduling problems (ALSP. The MMR-n was shown as a computational metaphor capable of optimizing the cultivar coefficients of CERES-Rice model, and at the same time, able to find solutions to TSP and ALSP. The MMR-n as a computational paradigm has a better computational wall clock time compared to when these three problems are solved individually by a single-molecule reactor (MMR-1.
Novel electrode structure in a DBD reactor applied to the degradation of phenol in aqueous solution
Mercado-Cabrera, Antonio; Peña-Eguiluz, Rosendo; López-Callejas, Régulo; Jaramillo-Sierra, Bethsabet; Valencia-Alvarado, Raúl; Rodríguez-Méndez, Benjamín; Muñoz-Castro, Arturo E.
2017-07-01
Phenol degradation experimental results are presented in a similar wastewater aqueous solution using a non-thermal plasma reactor in a coaxial dielectric barrier discharge. The novelty of the work is that one of the electrodes of the reactor has the shape of a hollow screw which shows an enhanced efficiency compared with a traditional smooth structure. The experimentation was carried out with gas mixtures of 90% Ar-10% O2, 80% Ar-20% O2 and 0% Ar-100% O2. After one hour of treatment the removal efficiency was 76%, 92%, and 97%, respectively, assessed with a gas chromatographic mass spectrometry technique. For both reactors used, the ozone concentration was measured. The screw electrode required less energy, for all gas mixtures, than the smooth electrode, to maintain the same ozone concentration. On the other hand, it was also observed that in both electrodes the electrical conductivity of the solution changed slightly from ˜0.0115 S m-1 up to ˜0.0430 S m-1 after one hour of treatment. The advantages of using the hollow screw electrode structure compared with the smooth electrode were: (1) lower typical power consumption, (2) the generation of a uniform plasma throughout the reactor benefiting the phenol degradation, (3) a relatively lower temperature of the aqueous solution during the process, and (4) the plasma generation length is larger.
International Nuclear Information System (INIS)
Souto, F.J.; Heger, A.S.
2001-01-01
To investigate the effects of radiolytic gas bubbles and thermal expansion on the steady-state operation of solution reactors at the power level required for the production of medical isotopes, a calculational model has been developed. To validate this model, including its principal hypotheses, specific experiments at the Los Alamos National Laboratory SHEBA uranyl fluoride solution reactor were conducted. The following sections describe radiolytic gas generation in solution reactors, the equations to estimate the fuel solution volume change due to radiolytic gas bubbles and thermal expansion, the experiments conducted at SHEBA, and the comparison of experimental results and model calculations. (author)
International Nuclear Information System (INIS)
Vasile, A.
2001-01-01
Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)
International Nuclear Information System (INIS)
Lee, Yoonhee; Cho, Nam Zin
2015-01-01
Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal
Slezak, Izabella H; Jasik-Slezak, Jolanta; Bilewicz-Wyrozumska, Teresa; Slezak, Andrzej
2006-01-01
On the basis of model equation describing the membrane potential delta psi(s) on concentration Rayleigh number (R(C)), mechanical pressure difference (deltaP), concentration polarization coefficient (zeta s) and ratio concentration of solutions separated by membrane (Ch/Cl), the characteristics delta psi(s) = f(Rc)(delta P, zeta s, Ch/Cl) for steady values of zeta s, R(C) and Ch/Cl in single-membrane system were calculated. In this system neutral and isotropic polymeric membrane oriented in horizontal plane, the non-homogeneous binary electrolytic solutions of various concentrations were separated. Nonhomogeneity of solutions is results from creations of the concentration boundary layers on both sides of the membrane. Calculations were made for the case where on a one side of the membrane aqueous solution of NaCl at steady concentration 10(-3) mol x l(-1) (Cl) was placed and on the other aqueous solutions of NaCl at concentrations from 10(-3) mol x l(-1) to 2 x 10(-2) mol x l(-1) (Ch). Their densities were greater than NaCl solution's at 10(-3) mol x l(-1). It was shown that membrane potential depends on hydrodynamic state of a complex concentration boundary layer-membrane-concentration boundary layer, what is controlled by deltaP, Ch/Cl, Rc and Zeta(s).
Cokoja, Mirza
2011-08-29
A plethora of methods have been developed over the years so that carbon dioxide can be used as a reactant in organic synthesis. Given the abundance of this compound, its utilization in synthetic chemistry, particularly on an industrial scale, is still at a rather low level. In the last 35 years, considerable research has been performed to find catalytic routes to transform CO 2 into carboxylic acids, esters, lactones, and polymers in an economic way. This Review presents an overview of the available homogeneous catalytic routes that use carbon dioxide as a C 1 carbon source for the synthesis of industrial products as well as fine chemicals. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
International Nuclear Information System (INIS)
Young, P.
2012-01-01
There has recently been a renewed interest throughout the world in small nuclear units for generating electricity and for other applications. A report by the World Nuclear Association discussing the advantages of small modular nuclear reactors (SMRs) over traditional nuclear reactor designs, states that ''modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction.'' Since the inception of nuclear power, the size of reactor units has grown from under 100 MWe to more than 1600 MWe. Today, due partly to the high capital cost of large power reactors and partly to the need to service small electricity grids, there is a move to develop smaller units. These may be built individually or as modules in a larger plant. SMRs are a good fit in markets where anticipated electricity demand is projected to increase incrementally, because SMRs could be built in series as needed. SMRs might be particularly attractive in countries that currently rely on diesel generators for producing electricity. Small reactors could make economic sense because of the high cost of diesel generation compared to the low marginal cost of producing electricity from nuclear energy. (Keeping in mind the initial investment costs and the need to establish a national regulatory program.) Some SMR designs are fabricated in a factory and then delivered to the site. This could be a solution for markets that lack the qualified engineers and skilled craft workers needed to construct large reactors on site. This paper will provide an overview of the types and attributes of SMRs in use or under development worldwide, describe the similarities and important differences between designs, discuss potential applications for SMRs, including baseload electricity generation, electricity generation for remote locations and areas with
Directory of Open Access Journals (Sweden)
Chifu E. N.
2009-10-01
Full Text Available In this article, we formulate solutions to Einstein's geometrical field equations derived using our new approach. Our field equations exterior and interior to the mass distribution have only one unknown function determined by the mass or pressure distribution. Our obtained solutions yield the unknown function as generalizations of Newton's gravitational scalar potential. Thus, our solution puts Einstein's geometrical theory of gravity on same footing with Newton's dynamical theory; with the dependence of the field on one and only one unknown function comparable to Newton's gravitational scalar potential. Our results in this article are of much significance as the Sun and planets in the solar system are known to be more precisely oblate spheroidal in geometry. The oblate spheroidal geometries of these bodies have effects on their gravitational fields and the motions of test particles and photons in these fields.
Reactor neutron converter into antineutrinos on base of lithium compounds and their solutions
International Nuclear Information System (INIS)
Lyutostanskij, Yu.S.; Lyashchuk, V.I.
1990-01-01
A study on the possibility of using various lithium compounds as substances for reactor neutron converter into antineutrinos, was made. It is concluded that heavy water LiOD, LiODxD 2 O, LiD solutions are the most promising ones. They provide for high efficiency with reduction of the required lithium mass (300-350 times maximum) as compared with pure lithium converter
International Nuclear Information System (INIS)
Khotylev, V.A.; Hoogenboom, J.E.
1996-01-01
The paper presents new techniques for the solution of the nuclear reactor equation in diffusion approximation, that has enhanced efficiency and stability. The code system based on the new technique solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three dimensional geometry with reduced CPU time as compared to similar code systems of previous generations if well-posed neutronics problems are considered. Automated detection of ill-posed problem and selection of the appropriate numerical method makes the new code system capable of yielding a correct solution for wider range of problems without user intervention. (author)
Energy Technology Data Exchange (ETDEWEB)
Khotylev, V.A.; Hoogenboom, J.E. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands)
1996-07-01
The paper presents new techniques for the solution of the nuclear reactor equation in diffusion approximation, that has enhanced efficiency and stability. The code system based on the new technique solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three dimensional geometry with reduced CPU time as compared to similar code systems of previous generations if well-posed neutronics problems are considered. Automated detection of ill-posed problem and selection of the appropriate numerical method makes the new code system capable of yielding a correct solution for wider range of problems without user intervention. (author)
Energy Technology Data Exchange (ETDEWEB)
Huittinen, N., E-mail: n.huittinen@hzdr.de [Helmholtz-Zentrum Dresden - Rossendorf, Institute of Resource Ecology, Bautzner Landstraße 400, 01328 Dresden (Germany); Arinicheva, Y. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), 52425 Jülich (Germany); Kowalski, P.M.; Vinograd, V.L. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), 52425 Jülich (Germany); JARA High-Performance Computing, Schinkelstraße 2, 52062 Aachen (Germany); Neumeier, S. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), 52425 Jülich (Germany); Bosbach, D. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), 52425 Jülich (Germany); JARA High-Performance Computing, Schinkelstraße 2, 52062 Aachen (Germany)
2017-04-01
Here we study the homogeneity of Eu{sup 3+}-doped La{sub 1-x}Gd{sub x}PO{sub 4} (x = 0, 0.11, 0.33, 0.55, 0.75, 0.92, 1) monazite-type solid solutions by a combination of Raman and time-resolved laser fluorescence spectroscopies (TRLFS) with complementary quasi-random structure-based atomistic modeling studies. For the intermediate La{sub 0.45}Gd{sub 0.55}PO{sub 4} composition we detected a significant broadening of the Raman bands corresponding to the lattice vibrations of the LnO{sub 9} polyhedron, indicating much stronger distortion of the lanthanide cation site than the PO{sub 4} tetrahedron. A distortion of the crystal lattice around the dopant site was also confirmed in our TRLFS measurements of Eu{sup 3+} doped samples, where both the half width (FWHM) of the excitation peaks and the {sup 7}F{sub 2}/{sup 7}F{sub 1} ratio derived from the emission spectra increase for intermediate solid-solution compositions. The observed variation in FWHM correlates well with the simulated distribution of Eu···O bond distances within the investigated monazites. The combined results imply that homogenous Eu{sup 3+}-doped La{sub 1-x}Gd{sub x}PO{sub 4} monazite-type solid solutions are formed over the entire composition range, which is of importance in the context of using these ceramics for immobilization of radionuclides. - Highlights: •Homogenous Eu{sup 3+}-doped La{sub 1-x}Gd{sub x}PO{sub 4} monazite-type solid solutions have been synthesized. •Solid solution formation is accompanied by slight distortion of the LnO{sub 9} polyhedron. •Raman and laser spectroscopic trends are observed within the monazite series. •Results are explained with atomistic simulations of Eu-O bond distance distribution.
Rao, Heng; Bonin, Julien; Robert, Marc
2017-11-23
An iron-substituted tetraphenyl porphyrin bearing positively charged trimethylammonio groups at the para position of each phenyl ring catalyzes the photoinduced conversion of CO 2 . This complex is water soluble and acts as a molecular catalyst to selectively reduce CO 2 into CO under visible-light irradiation in aqueous solutions (acetonitrile/water=1:9 v/v) with the assistance of purpurin, a simple organic photosensitizer. CO is produced with a catalytic selectivity of 95 % and turnover number up to 120, illustrating the possibility of photocatalyzing the reduction of CO 2 in aqueous solution by using visible light, a simple organic sensitizer coupled to an amine as a sacrificial electron donor, and an earth-abundant metal-based molecular catalyst. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.
DEFF Research Database (Denmark)
Johannesson, Björn
2009-01-01
Results from a systematic continuum mixture theory will be used to establish the governing equations for ionic diffusion and chemical reactions in the pore solution of a porous material subjected to moisture transport. The theory in use is the hybrid mixture theory (HMT), which in its general form......’s law of diffusion and the generalized Darcy’s law will be used together with derived constitutive equations for chemical reactions within phases. The mass balance equations for the constituents and the phases together with the constitutive equations gives the coupled set of non-linear differential...... general description of chemical reactions among constituents is described. The Petrov – Galerkin approach are used in favour of the standard Galerkin weighting in order to improve the solution when the convective part of the problem is dominant. A modified type of Newton – Raphson scheme is derived...
Energy Technology Data Exchange (ETDEWEB)
El Ganaoui, K
2006-09-15
In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)
International Nuclear Information System (INIS)
Billard, Francois; Lavie, Jean-Marie
1964-10-01
Within the frame of the study of radiological risks associated with a reactor accident in order to define the required responses, this study comprises, on the one hand, an analysis of the different accident types in order to select typical accidents, and on the other hand, a site-based analysis to define the maximum admissible radioactivity release for a given site. The determination of minimum required coefficient of risk reduction results from a compromise between the choice of reactor configuration type and the efficiency of purification devices, while taking into account minimum characteristics of the enclosure mechanical strength, local release conditions, and nature of gaseous effluents to be processed. After a review of available containment techniques, the author applies this analysis method to the different French reactor types. He gives a brief description of adopted solutions for the most typical French reactors in terms of characteristics of venting and filtration devices. As data quality is a crucial requirement, the author outlines the need for further studies regarding fission product emission and transfer, the purification of gaseous effluents and their diffusion in the atmosphere [fr
International Nuclear Information System (INIS)
Goncalves, Glenio A.; Orengo, Gilberto; Vilhena, Marco Tullio M.B. de; Graca, Claudio O.
2002-01-01
In this work we present the LTS N solution of the adjoint transport equation for an arbitrary source, testing the aptness of this analytical solution for high order of quadrature in transport problems and comparing some preliminary results with the ANISN computations in a homogeneneous slab geometry. In order to do that we apply the new formulation for the LTS N method based on the invariance projection property, becoming possible to handle problems with arbitrary sources and demanding high order of quadrature or deep penetration. This new approach for the LTS N method is important both for direct and adjoint transport calculations and its development was inspired by the necessity of using generalized adjoint sources for important calculations. Although the mathematical convergence has been proved for an arbitrary source, when the quadrature order or deep penetration is required the LTS N method presents computational overflow even for simple sources (sin, cos, exp, polynomial). With the new formulation we eliminate this drawback and in this work we report the numerical simulations testing the new approach
Homogenization approach in engineering
International Nuclear Information System (INIS)
Babuska, I.
1975-10-01
Homogenization is an approach which studies the macrobehavior of a medium by its microproperties. Problems with a microstructure play an essential role in such fields as mechanics, chemistry, physics, and reactor engineering. Attention is concentrated on a simple specific model problem to illustrate results and problems typical of the homogenization approach. Only the diffusion problem is treated here, but some statements are made about the elasticity of composite materials. The differential equation is solved for linear cases with and without boundaries and for the nonlinear case. 3 figures, 1 table
International Nuclear Information System (INIS)
Calabrese, C.R.; Grant, C.R.
1990-01-01
This work presents comparisons between measured fluxes obtained by activation of Manganese foils in the light water, enriched uranium research pool reactor RA-2 MTR (Materials Testing Reactors) fuel element) and fluxes calculated by the finite element method FEM using DELFIN code, and describes the heterogeneus finite elements by a set of solutions of the transport equations for several different configurations obtained using the collision probability code HUEMUL. The agreement between calculated and measured fluxes is good, and the advantage of using FEM is showed because to obtain the flux distribution with same detail using an usual diffusion calculation it would be necessary 12000 mesh points against the 2000 points that FEM uses, hence the processing time is reduced in a factor ten. An interesting alternative to use in MTR fuel management is presented. (Author) [es
International Nuclear Information System (INIS)
Shimizu, Yoshiaki
1988-01-01
Due to the simplicity and effectiveness, linear program has been popular in the actual optimization in various fields. In the previous study, the uncertainty involved in the model at the different stage of optimization was dealt with by post-optimizing analysis. But it often becomes insufficient to make a decision how to deal with an uncertain system especially suffering large parameter deviation. Recently in the field of processing systems, it is desired to obtain a flexible solution which can present the counterplan to a deviating system from a practical viewpoint. The scope of this preliminary note presents how to apply a methodology development to obtain the flexible solution of a linear program. For this purpose, a simple example associated with nuclear reactor decommissioning is shown. The problem to maximize a system performance given as an objective function under the constraint of the static behavior of the system is considered, and the flexible solution is determined. In Japan, the decommissioning of commercial nuclear power plants will being in near future, and the study using the retired research reactor JPDR is in progress. The planning of decontamination and the reuse of wastes is taken as the example. (Kako, I.)
Energy Technology Data Exchange (ETDEWEB)
Petersen, Claudio Z. [Universidade Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Bodmann, Bardo E.J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia Mecanica; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro, Nova Friburgo, RJ (Brazil). Inst. Politecnico
2014-12-15
In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i.e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.
Temporary solutions for a conservative estimation of void reactivity insertion in CANDU reactor
International Nuclear Information System (INIS)
Dumitrache, I.
1997-01-01
One of the most difficult task of the CANDU Reactor Physics Analysis is related to the correct treatment of the deviations from the reference coolant properties. The most significant problem is the reactivity inserted by a given coolant density variation. From the practical Nuclear Safety Analysis point of view, the solution must be not only conservative, but also adaptable to the current chain of codes utilized for accident simulation. The first set of experimental data was obtained by AECL many years ago. The fuel was fresh, clean and cold. Some of the currently used computer codes offer accurate predictions of the measured void reactivities. Unfortunately, the existing experimental data do not cover and are not significant for the burned CANDU fuel. A specific benchmark problem was suggested by the Institute for Nuclear Research (ICN) Pitesti. The problem was analysed and slightly modified during an IAEA Vienna RCM (Research Coordinating Meeting), Buenos Aires, 1990. Afterwards, the problem was independently solved in several countries, interested by the CANDU reactor. The results were presented and analysed at the Bombay RCM, 1992. It was clear that the interval defined by the code predictions is much too broad. New experimental data are necessary. They must cover the fuel isotopic composition specific for the burned CANDU fuel. The work is in progress at the Chalk River Laboratory. Temporary solutions have been analysed at the ICN Pitesti. The first aim was to identify the reactivity numerical values that are conservative, but not too inaccurate. The WIMS code predictions have been compared against other estimations, including the Monte-Carlo based ones. The second aim was to force the currently used code, PPV, to offer cell cross sections that are correct from the Reactor Physics point of view, and compatible with the imposed reactivity. Physical and mathematical procedures were proposed and evaluated. An additional solution was also taken into account: to
International Nuclear Information System (INIS)
Cai, Li
2014-01-01
In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core
Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor
Directory of Open Access Journals (Sweden)
Nowak Tomasz Karol
2014-06-01
Full Text Available This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.
Engineering solutions for components facing the plasma in experimental power reactors
International Nuclear Information System (INIS)
Casini, G.; Farfaletti-Casali, F.
1985-01-01
A review of the engineering problems related to the structures in front of the plasma of experimental Tokamak-type reactors is made. Attention is focused on the so-named ''first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system, in particular for the case of the single-null poloidal divertor. Even if the uncertainties related to the plasma-wall interaction are stil relevant, some engineering solutions which look manageable are identified and described. (orig.)
Features of the Numerical Solution of Thermal Destruction Fuel Pins Problems in the Fast Reactor
Usov, E. V.; Butov, A. A.; Klimonov, I. A.; Chuhno, V. I.; Nikolaenko, A. V.; Zhdanov, V. S.; Pribaturin, N. A.; Strizhov, V. F.
2017-11-01
In this paper the description of the basic equations which can be used for calculation of melting of fuel and cladding of the fast reactor, moving of the melt on a fuel pin surface and its solidification is presented. The special attention is given speed of calculation algorithms and fidelity of the phenomena which are observed at a stage of severe accidents in fast reactors. For check of working capacity of initial models, numerical calculations of Stefan-type problems on front movement of melting/solidification in cylindrical geometry are presented. Comparison with the solutions received by known analytical methods is executed. For validation of the numerical realization of calculation algorithms the analysis is carried out and experiments in which melting of the model fuel pins of fast reactors was studied are chosen. On the basis of the chosen experiments calculation schemes taking into account initial and boundary conditions are prepared and modeling is performed. Modeling results are shown in the present paper. Estimation of calculation error of the basic physical parameters is done by results of the modeling and conclusions are drawn on a correctness of algorithms operation.
Directory of Open Access Journals (Sweden)
Wenzhen Chen
2013-01-01
Full Text Available The singularly perturbed method (SPM is proposed to obtain the analytical solution for the delayed supercritical process of nuclear reactor with temperature feedback and small step reactivity inserted. The relation between the reactivity and time is derived. Also, the neutron density (or power and the average density of delayed neutron precursors as the function of reactivity are presented. The variations of neutron density (or power and temperature with time are calculated and plotted and compared with those by accurate solution and other analytical methods. It is shown that the results by the SPM are valid and accurate in the large range and the SPM is simpler than those in the previous literature.
Prestart-up hydrogen peroxide solution washing of NPP unit with the RBMK-type reactor
International Nuclear Information System (INIS)
Gruzdev, N.I.; Man'kina, N.N.; Al'tshuller, M.A.
1979-01-01
Presented are the results of industrial hydrogen peroxide solution washing of condensating-feed system conducted on the second unit of the Kursk NPP. Duration of the washing constituted 8 hours. The hydrogen peroxide concentration during first 4 hours was 10-20 mg/kg at a flow rate of 260 m 2 /h, during the following 4 hours it constituted 2-5 mg/kg at a flow rate of 1000 m 3 /h. It is found out that prestart-up hydrogen peroxide washing of NPP power units with the RBMK-type reactor permits: to simplify essentially the technology and scheme of washing process; to reduce a flow rate of desalt washing water; to except environmental contamination with washing solutions and reagents being neutralized; to reduce the time of washing process; to reduce the time necessary for the achievement of reference water condition factors, and to increase the unit reliability and to improve a radiation situation
International Nuclear Information System (INIS)
Sanchez, R.; Ragusa, J.; Santandrea, S.
2004-01-01
The problem of the determination of a homogeneous reflector that preserves a set of prescribed albedo is considered. Duality is used for a direct estimation of the derivatives needed in the iterative calculation of the optimal homogeneous cross sections. The calculation is based on the preservation of collapsed multigroup albedo obtained from detailed reference calculations and depends on the low-order operator used for core calculations. In this work we analyze diffusion and transport as low-order operators and argue that the P 0 transfers are the best choice for the unknown cross sections to be adjusted. Numerical results illustrate the new approach for SP N core calculations. (Author)
Energy Technology Data Exchange (ETDEWEB)
Palma, Daniel A.P. [CEFET QUIMICA de Nilopolis/RJ, 21941-914 Rio de Janeiro (Brazil)], E-mail: agoncalves@con.ufrj.br; Martinez, Aquilino S.; Goncalves, Alessandro C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)
2009-09-15
The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.
International Nuclear Information System (INIS)
Palma, Daniel A.P.; Martinez, Aquilino S.; Goncalves, Alessandro C.
2009-01-01
The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.
International Nuclear Information System (INIS)
El-Morshedy, Salah El-Din
2010-01-01
Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Nogueira, K.R.B.; Nascimento, C.A.O.; Guardani, R.; Teixeira, A.C.S.C. [University of Sao Paulo, Chemical Engineering Department, Sao Paulo (Brazil)
2012-12-15
Solar reactors can be attractive in photodegradation processes due to lower electrical energy demand. The performance of a solar reactor for two flow configurations, i.e., plug flow and mixed flow, is compared based on experimental results with a pilot-scale solar reactor. Aqueous solutions of phenol were used as a model for industrial wastewater containing organic contaminants. Batch experiments were carried out under clear sky, resulting in removal rates in the range of 96-100 %. The dissolved organic carbon removal rate was simulated by an empirical model based on neural networks, which was adjusted to the experimental data, resulting in a correlation coefficient of 0.9856. This approach enabled to estimate effects of process variables which could not be evaluated from the experiments. Simulations with different reactor configurations indicated relevant aspects for the design of solar reactors. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)
International Nuclear Information System (INIS)
Cui, D.; Eriksen, T.E.
1996-03-01
The redox chemistry of Technetium and Neptunium in deep groundwater systems has been studied under well controlled conditions in laboratory experiments. The measured redox potentials in anoxic deep groundwater systems are consistent with redox reactions between Fe(II) in solution and hydrous Fe(III)-oxide phases. The fracture filling material and groundwater in transmissive fractures in bedrock constitute two different compartments in the groundwater system and experiments were therefore carried out in homogeneous Fe(II) containing solutions and in heterogeneous mixtures of solution with Fe(II) containing solid mineral phases. Reduction of the strongly sorbing neptunyl cation (NpO 2 + ) and the slightly sorbing pertechnetate anion (TcO 4 - ) by Fe(II) in solution was found to proceed very slowly, if at all, in reaction vessels with hydrophobic inner surfaces. However, in the heterogeneous systems we observed surface mediated reduction to the slightly soluble ( -8 mol*dm -3 ) tetravalent (hydr)oxides TcO 2 *nH 2 O (=Tc(OH) 4 ) and NpO 2 *nH 2 O (=Np(OH) 4 ) by Fe(II) sorbed on quartz,precipitated Fe(OH) 2 (s), Fe(II)CO 3 (s) and Fe(II) bearing minerals such as magnetite, hornblende and Fe(II)-chlorite. It is concluded that surface mediated redox-reactions will be the most effective pathway for the reduction of Tc(VII) and Np(V) in deep groundwater systems. On exposure of the surface-precipitated tetravalent (hydr)oxides to air saturated groundwater solutions the oxidative dissolution was found to be a very slow process and high concentration of hydrogen peroxide was required for oxidative dissolution. The slow rate of oxidative dissolution is most probably due to kinetic suppression of the reactions between dissolved oxygen and the precipitated (hydr)oxides. The kinetic suppression is caused by competing redox reactions at the surface of the Fe(II)-bearing minerals which consumes the dissolved oxygen. 30 refs, 22 figs
Energy Technology Data Exchange (ETDEWEB)
Cui, D.; Eriksen, T.E. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Chemistry
1996-03-01
The redox chemistry of Technetium and Neptunium in deep groundwater systems has been studied under well controlled conditions in laboratory experiments. The measured redox potentials in anoxic deep groundwater systems are consistent with redox reactions between Fe(II) in solution and hydrous Fe(III)-oxide phases. The fracture filling material and groundwater in transmissive fractures in bedrock constitute two different compartments in the groundwater system and experiments were therefore carried out in homogeneous Fe(II) containing solutions and in heterogeneous mixtures of solution with Fe(II) containing solid mineral phases. Reduction of the strongly sorbing neptunyl cation (NpO{sub 2}{sup +}) and the slightly sorbing pertechnetate anion (TcO{sub 4}{sup -}) by Fe(II) in solution was found to proceed very slowly, if at all, in reaction vessels with hydrophobic inner surfaces. However, in the heterogeneous systems we observed surface mediated reduction to the slightly soluble (<10{sub -8} mol*dm{sup -3}) tetravalent (hydr)oxides TcO{sub 2}*nH{sub 2}O (=Tc(OH){sub 4}) and NpO{sub 2}*nH{sub 2}O (=Np(OH){sub 4}) by Fe(II) sorbed on quartz,precipitated Fe(OH){sub 2}(s), Fe(II)CO{sub 3}(s) and Fe(II) bearing minerals such as magnetite, hornblende and Fe(II)-chlorite. It is concluded that surface mediated redox-reactions will be the most effective pathway for the reduction of Tc(VII) and Np(V) in deep groundwater systems. On exposure of the surface-precipitated tetravalent (hydr)oxides to air saturated groundwater solutions the oxidative dissolution was found to be a very slow process and high concentration of hydrogen peroxide was required for oxidative dissolution. The slow rate of oxidative dissolution is most probably due to kinetic suppression of the reactions between dissolved oxygen and the precipitated (hydr)oxides. The kinetic suppression is caused by competing redox reactions at the surface of the Fe(II)-bearing minerals which consumes the dissolved oxygen.
Energy Technology Data Exchange (ETDEWEB)
Sanchez, R.; Ragusa, J.; Santandrea, S. [Commissariat a l' Energie Atomique, Direction de l' Energie Nucleaire, Service d' Etudes de Reacteurs et de Modelisation Avancee, CEA de Saclay, DM2S/SERMA 91 191 Gif-sur-Yvette cedex (France)]. e-mail: richard.sanchez@cea.fr
2004-07-01
The problem of the determination of a homogeneous reflector that preserves a set of prescribed albedo is considered. Duality is used for a direct estimation of the derivatives needed in the iterative calculation of the optimal homogeneous cross sections. The calculation is based on the preservation of collapsed multigroup albedo obtained from detailed reference calculations and depends on the low-order operator used for core calculations. In this work we analyze diffusion and transport as low-order operators and argue that the P{sub 0} transfers are the best choice for the unknown cross sections to be adjusted. Numerical results illustrate the new approach for SP{sub N} core calculations. (Author)
Reactor core in FBR type reactor
International Nuclear Information System (INIS)
Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.
1989-01-01
In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)
Integrated, Reactor Relevant Solutions for Lower Hybrid Range of Frequencies Actuators
Shiraiwa, S.; Bonoli, P. T.; Lin, Y.; Wallace, G. M.; Wukitch, S. J.
2017-10-01
RF (radiofrequency) actuators with high system efficiency (wall-plug to plasma) and ability for continuous operation have long be recognized as essential tools for realizing a steady state tokamak. A number of physics and technological challenges to utilization remain including current drive efficiency and location, efficient coupling, and impurity contamination. In a reactor environment, plasma material interaction (PMI) issues associated with coupling structures are similar to the first wall and have been identified as a potential show-stopper. High field side (HFS) launch of LHRF power represents an integrated solution that both improves core wave physics and mitigates PMI/coupling issues. For HFS LHRF, wave penetration is vastly improves because wave accessibility scales as 1/B allowing for launching the wave at lower n|| (parallel refractive index). The lower n|| penetrate to higher electron temperature resulting in higher current drive efficiency (1/n||2). HFS RF launch also provides for a means to dramatically improve launcher robustness in a reactor environment. On the HFS, the SOL is quiescent; local density profile is steep and controlled through magnetic shape; fast particle, neutron, turbulent heat and particle fluxes are eliminated or minim Work supported by the U.S. DoE, Office of Science, Office of Fusion Energy Sciences, User Facility Alcator C-Mod under DE-FC02-99ER54512 and US DoE Contract No. DE-FC02-01ER54648 under a Scientific Discovery through Advanced Computing Initiative.
Abstract of programs for nuclear reactor calculation and kinetic equations solution
International Nuclear Information System (INIS)
Marakazov, A.A.
1977-01-01
The collection includes about 50 annotations of programmes,developed in the Kurchatov Atomic Energy Institute in 1971-1976. The programmes are intended for calculating the neutron flux, for solving systems of multigroup equations in P 3 approximation, for calculating the reactor cell, for analysing the system stability, breeding ratio etc. The programme annotations are compiled according to the following diagram: 1.Programme title. 2.Computer type. 3.Physical problem. 4.Solution method. 5.Calculation limitations. 6.Characteristic computer time. 7.Programme characteristic features. 8.Bound programmes. 9.Programme state. 10.Literature allusions in the programme. 11.Required memory resourses. 12.Programming language. 13.Operation system. 14.Names of authors and place of programme adjusting
International Nuclear Information System (INIS)
Lucas, M.
1988-01-01
Measurements were made of I/sub 2/ formed when aqueous cesium iodide (CsI) solutions were exposed to two temperatures, 43 and 95 0 C, with irradiation. Iodine partition coefficients were obtained from the experiments. The parameters varied were dose, CsI concentration, and Cs/sub 2/CO/sub 3/ concentration, in the presence of air-carbon dioxide and air-carbon dioxide-hydrogen mixtures, to provide information to calculate the form in which iodine released from fuel as CsI in a reactor accident might reach the environment. In a series of experiments, a two-compartment cell was used to trap the gaseous iodine produced. In this case, it was found that the quantity of gaseous iodine produced increased approximately linearly with the dose (at the dose rate used)
Stainless steels in boiling water reactors. Corrosion problems and possible solutions
International Nuclear Information System (INIS)
Combrade, P.; Desestret, A.; Leroy, F.; Coriou, H.
1977-01-01
In boiling water reactors, the heat-carrying water may have an up to 0.1 or even 0.2 ppm oxygen content, which can make it highly agressive at operating temperature for stainless steels subject to high physical stresses. Several metallurgical solutions can be considered, and in particular the use of stainless steels having a mixed austenitic-ferritic structure or of standard austenitic steels (18.10 or 18.10 Mo, such as AISI 304 and 316) with carefully controlled carbon and alloy element contents. The behavior of these steels during prolonged tests in water at 288 0 C with a 30 and even 100 ppm oxygen content turned out to be quite satisfactory [fr
Benchmarking monthly homogenization algorithms
Venema, V. K. C.; Mestre, O.; Aguilar, E.; Auer, I.; Guijarro, J. A.; Domonkos, P.; Vertacnik, G.; Szentimrey, T.; Stepanek, P.; Zahradnicek, P.; Viarre, J.; Müller-Westermeier, G.; Lakatos, M.; Williams, C. N.; Menne, M.; Lindau, R.; Rasol, D.; Rustemeier, E.; Kolokythas, K.; Marinova, T.; Andresen, L.; Acquaotta, F.; Fratianni, S.; Cheval, S.; Klancar, M.; Brunetti, M.; Gruber, C.; Prohom Duran, M.; Likso, T.; Esteban, P.; Brandsma, T.
2011-08-01
The COST (European Cooperation in Science and Technology) Action ES0601: Advances in homogenization methods of climate series: an integrated approach (HOME) has executed a blind intercomparison and validation study for monthly homogenization algorithms. Time series of monthly temperature and precipitation were evaluated because of their importance for climate studies and because they represent two important types of statistics (additive and multiplicative). The algorithms were validated against a realistic benchmark dataset. The benchmark contains real inhomogeneous data as well as simulated data with inserted inhomogeneities. Random break-type inhomogeneities were added to the simulated datasets modeled as a Poisson process with normally distributed breakpoint sizes. To approximate real world conditions, breaks were introduced that occur simultaneously in multiple station series within a simulated network of station data. The simulated time series also contained outliers, missing data periods and local station trends. Further, a stochastic nonlinear global (network-wide) trend was added. Participants provided 25 separate homogenized contributions as part of the blind study as well as 22 additional solutions submitted after the details of the imposed inhomogeneities were revealed. These homogenized datasets were assessed by a number of performance metrics including (i) the centered root mean square error relative to the true homogeneous value at various averaging scales, (ii) the error in linear trend estimates and (iii) traditional contingency skill scores. The metrics were computed both using the individual station series as well as the network average regional series. The performance of the contributions depends significantly on the error metric considered. Contingency scores by themselves are not very informative. Although relative homogenization algorithms typically improve the homogeneity of temperature data, only the best ones improve precipitation data
Electro-magnetostatic homogenization of bianisotropic metamaterials
Fietz, Chris
2012-01-01
We apply the method of asymptotic homogenization to metamaterials with microscopically bianisotropic inclusions to calculate a full set of constitutive parameters in the long wavelength limit. Two different implementations of electromagnetic asymptotic homogenization are presented. We test the homogenization procedure on two different metamaterial examples. Finally, the analytical solution for long wavelength homogenization of a one dimensional metamaterial with microscopically bi-isotropic i...
International Nuclear Information System (INIS)
Toyama, Masahiro; Kasai, Shigeo.
1978-01-01
Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)
Energy Technology Data Exchange (ETDEWEB)
Wahab, Mohamed Ali, E-mail: waheb_med@yahoo.fr [University of Carthage, Water Research and Technologies Centre (CERTE), Wastewater Treatment and Recycling Laboratory, B.P. 273, 8020 Soliman (Tunisia); Hassine, Rafik Ben [International Environmental Green Technology (IGET) (Tunisia); Jellali, Salah, E-mail: salah.jallali@certe.rnrt.tn [University of Carthage, Water Research and Technologies Centre (CERTE), Wastewater Treatment and Recycling Laboratory, B.P. 273, 8020 Soliman (Tunisia)
2011-05-15
The present study aims to develop a new potentially low-cost, sustainable treatment approach to soluble inorganic phosphorus removal from synthetic solutions and secondary wastewater effluents in which a plant waste (Posidonia oceanica fiber: POF) is used for further agronomic benefit. Dynamic flow tests using a continuous stirred tank reactor (CSTR) were carried out to study the effect of initial concentration of phosphorus, amount of adsorbent, feeding flow rate and anions competition. The experimental results showed that the removal efficiency of phosphorus from synthetic solutions is about 80% for 10 g L{sup -1} of POF. In addition, the variation of the initial concentration of phosphorus from 8 to 50 mg L{sup -1} increased the adsorption capacity from 0.99 to 3.03 mg g{sup -1}. The use of secondary treated wastewater showed the presence of competition phenomenon between phosphorus and sulphate which could be overcoming with increasing the sorptive surface area and providing more adsorption sites when increasing the adsorbent dosage of POF. Compared with columns studies, this novel CSTR system showed more advantages for the removal of soluble phosphorus as a tertiary treatment of urban secondary effluents with more adsorption efficiency and capacity, in addition to the prospect use of saturated POF with nutriment as fertilizer and compost.
International Nuclear Information System (INIS)
Wahab, Mohamed Ali; Hassine, Rafik Ben; Jellali, Salah
2011-01-01
The present study aims to develop a new potentially low-cost, sustainable treatment approach to soluble inorganic phosphorus removal from synthetic solutions and secondary wastewater effluents in which a plant waste (Posidonia oceanica fiber: POF) is used for further agronomic benefit. Dynamic flow tests using a continuous stirred tank reactor (CSTR) were carried out to study the effect of initial concentration of phosphorus, amount of adsorbent, feeding flow rate and anions competition. The experimental results showed that the removal efficiency of phosphorus from synthetic solutions is about 80% for 10 g L -1 of POF. In addition, the variation of the initial concentration of phosphorus from 8 to 50 mg L -1 increased the adsorption capacity from 0.99 to 3.03 mg g -1 . The use of secondary treated wastewater showed the presence of competition phenomenon between phosphorus and sulphate which could be overcoming with increasing the sorptive surface area and providing more adsorption sites when increasing the adsorbent dosage of POF. Compared with columns studies, this novel CSTR system showed more advantages for the removal of soluble phosphorus as a tertiary treatment of urban secondary effluents with more adsorption efficiency and capacity, in addition to the prospect use of saturated POF with nutriment as fertilizer and compost.
International Nuclear Information System (INIS)
Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.
1979-01-01
Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)
Littel, Rob J.; Versteeg, Geert F.; Swaaij, Wim P.M. van
1992-01-01
Absorption experiments of COS into aqueous solutions of MDEA and DEMEA at 303 K have been carried out in a stirred cell reactor. An absorption model, based on Higbie’s penetration theory, has been developed and applied to interpret the absorption experiments, using the kinetic data obtained in part
International Nuclear Information System (INIS)
Jachic, J.
1987-01-01
The azimuthal neutronic flux distribution in the control ring region for a low power fast reactor is simulated using a plate and rectangular smash models for one dimensional calculations under periodic boundary conditions in the frontier. (E.G.) [pt
The European Pressurized Water Reactor. A safe and competitive solution for future energy needs
International Nuclear Information System (INIS)
Leverenz, R.; Gerhard, L.; Goebel, A.
2004-01-01
The European Pressurized Water Reactor - the EPR - is a PWR in the 1600 MW class. Its design is based on experience feedback from several thousand reactors x years of light water reactor operation worldwide, primarily those incorporating the most recent technologies: the French N4 and the German KONVOI reactors. It is an evolutionary design that ensures continuity in the mastery of PWR technology, minimizing the risk for the customer. (author)
Directory of Open Access Journals (Sweden)
Wei Guan
2017-01-01
Full Text Available In this experiment, the porous calcium silicate hydrates (P-CSHs were prepared via a hydrothermal method and then modified by polyethylene glycol (PEG. The modified P-CSHs combined with an internal recycle reactor could successfully recover the phosphorus from electroplating wastewater. The modified P-CSHs were characterized by X-ray diffraction (XRD, N2 adsorption-desorption isotherms, and Fourier transform infrared spectroscopy (FT-IR. After compared with different samples, the modified P-CSHs-PEG2000 sample had larger specific surface area of 87.48 m2/g and higher pore volume of 0.33 cm3/g, indicating a high capacity for phosphorus recovery. In the process of phosphorus recovery, the pH value of solution was increased to 9.5, which would enhance the recovery efficiency of phosphorus. The dissolution rate of Ca2+ from P-CSH-PEG2000 was fast, which was favorable for phosphorus precipitation and phosphorus recovery. The effects of initial concentration of phosphorus, P-CSHs-PEG2000 dosage, and stirring speed on phosphorus recovery were analyzed, so the optimal operation conditions for phosphorus recovery were obtained. The deposition was analyzed by XRD, N2 adsorption-desorption, and SEM techniques; it was indicated that the pore volume and surface area of the P-CSHs-PEG2000 were significantly reduced, and the deposition on the surface of P-CSHs-PEG2000 was hydroxyapatite.
International Nuclear Information System (INIS)
1979-06-01
This paper deals with the heavy water reactor, which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources not wishing to be dependent on the import of enrichment and other fuel cycle services which, in addition, would draw on the foreign exchange reserves. Pressurized heavy water reactors have been designed and built according to two somewhat different versions. While the Canadian CANDU-PHWR concept uses pressure tubes in a nearly unpressurized moderator tank (calandria), the German development line takes advantage of the established and well proven LWR technology, and, thus, uses a pressure vessel design where coolant channels and the surrounding moderator are held at equal pressure. This pressure vessel type heavy water reactor which has been built on a commercial demonstration plant level at ATUCHA in Argentina is described in a companion paper where also a conceptual design for a 685 MWsub(e) PHWR is discussed
Homogenization methods for heterogeneous assemblies
International Nuclear Information System (INIS)
Wagner, M.R.
1980-01-01
The third session of the IAEA Technical Committee Meeting is concerned with the problem of homogenization of heterogeneous assemblies. Six papers will be presented on the theory of homogenization and on practical procedures for deriving homogenized group cross sections and diffusion coefficients. That the problem of finding so-called ''equivalent'' diffusion theory parameters for the use in global reactor calculations is of great practical importance. In spite of this, it is fair to say that the present state of the theory of second homogenization is far from being satisfactory. In fact, there is not even a uniquely accepted approach to the problem of deriving equivalent group diffusion parameters. Common agreement exists only about the fact that the conventional flux-weighting technique provides only a first approximation, which might lead to acceptable results in certain cases, but certainly does not guarantee the basic requirement of conservation of reaction rates
International Nuclear Information System (INIS)
Baskar, S.; Jose, M.T.; Baskaran, R.; Venkatraman, B.
2018-01-01
The diluted virgin solutions (both aqueous and organic) and aqueous analytical waste generated from experimental analysis of process solutions, pertaining to Fast Breeder Test Reactor (FBTR) and Prototype Fast Breeder Reactor (PFBR), in glove boxes of active analytical Laboratory (AAL) are pumped back to the process cells through a pipe in pipe arrangement. There are 6 transfer lines (Length 15-32 m), 2 for each type of transfer. The transfer lines passes through the area inside the AAL and also the operating area. Hence it is required to compute the necessary radial shielding requirement around the lines to limit the dose rates in both the areas to the permissible values as per the regulatory requirement
International Nuclear Information System (INIS)
Moura Neto, C. de; Nair, R.P.K.
1979-08-01
The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt
International Nuclear Information System (INIS)
Alam, S.B.; Lindley, B.A.; Parks, G.T.
2015-01-01
In this reactor physics study, we attempt to design a civil marine reactor core that can operate over a 10 effective-full-power-years life at 333 MWth using ThUO 2 and all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements, optimizing: subassembly and core geometry; fuel enrichment; burnable and moveable poison design; and whole-core loading patterns. We compare designs with a 14% fissile loading for ThUO 2 and all-UO 2 fuel in 13*13 assemblies with ZrB 2 integral fuel burnable absorber pins for reactivity control. Taking advantage of self-shielding effects, the ThUO 2 option shows greater promise in the final burnable poison design while maintaining low, stable reactivity with minimal burnup penalty. For the final poisoning design with ZrB 2 , ThUO 2 contributes 2.5% more initial reactivity suppression, although the all-UO 2 design exhibits lower reactivity swing. All the candidate materials show greater rod worth for the ThUO 2 design. For both fuels, B 4 C has the highest reactivity worth, providing 10% higher control rod worth for ThUO 2 fuel than all-UO 2 . Finally, optimized assemblies were loaded into a 3D reactor model in PANTHER. The PANTHER results show that after 10 years, the core is on the border of criticality, confirming the fissile loading is well-designed. (authors)
Magnetite synthesis from ferrous iron solution at pH 6.8 in a continuous stirred tank reactor.
Mos, Yvonne M; Zorzano, Karin Bertens; Buisman, Cees J N; Weijma, Jan
2018-04-01
Partial oxidation of defined Fe 2+ solutions is a well-known method for magnetite synthesis in batch systems. The partial oxidation method could serve as basis for an iron removal process in drinking water production, yielding magnetite (Fe 3 O 4 ) as a compact and valuable product. As a first step toward such a process, a series of experiments was carried out, in which magnetite was synthesized from an Fe 2+ solution in a 2 L continuous stirred tank reactor (CSTR) at atmospheric pressure and 32 °C. In four experiments, elevating the pH from an initial value of 5.5 or 6.0 to a final value of 6.8, 7.0 or 7.5 caused green rust to form, eventually leading to magnetite. Formation of NH 4 + in the reactor indicated that NO 3 - and subsequently NO 2 - served as the oxidant. However, mass flow analysis revealed an influx of O 2 to the reactor. In a subsequent experiment, magnetite formation was achieved in the absence of added nitrate. In another experiment, seeding with magnetite particles led to additional magnetite precipitation without the need for a pH elevation step. Our results show, for the first time, that continuous magnetite formation from an Fe 2+ solution is possible under mild conditions, without the need for extensive addition of chemicals.
Continuous formation of N-chloro-N,N-dialkylamine solutions in well-mixed meso-scale flow reactors
Jolley, Katherine E
2015-01-01
Summary The continuous flow synthesis of a range of organic solutions of N,N-dialkyl-N-chloramines is described using either a bespoke meso-scale tubular reactor with static mixers or a continuous stirred tank reactor. Both reactors promote the efficient mixing of a biphasic solution of N,N-dialkylamine in organic solvent, and aqueous sodium hypochlorite to achieve near quantitative conversions, in 72–100% in situ yields, and useful productivities of around 0.05 mol/h with residence times from 3 to 20 minutes. Initial calorimetric studies have been carried out to inform on reaction exotherms, rates and safe operation. Amines which partition mainly in the organic phase require longer reaction times, provided by the CSTR, to compensate for low mass transfer rates in the biphasic system. The green metrics of the reaction have been assessed and compared to existing procedures and have shown the continuous process is improved over previous procedures. The organic solutions of N,N-dialkyl-N-chloramines produced continuously will enable their use in tandem flow reactions with a range of nucleophilic substrates. PMID:26734089
Continuous formation of N-chloro-N,N-dialkylamine solutions in well-mixed meso-scale flow reactors
Directory of Open Access Journals (Sweden)
A. John Blacker
2015-12-01
Full Text Available The continuous flow synthesis of a range of organic solutions of N,N-dialkyl-N-chloramines is described using either a bespoke meso-scale tubular reactor with static mixers or a continuous stirred tank reactor. Both reactors promote the efficient mixing of a biphasic solution of N,N-dialkylamine in organic solvent, and aqueous sodium hypochlorite to achieve near quantitative conversions, in 72–100% in situ yields, and useful productivities of around 0.05 mol/h with residence times from 3 to 20 minutes. Initial calorimetric studies have been carried out to inform on reaction exotherms, rates and safe operation. Amines which partition mainly in the organic phase require longer reaction times, provided by the CSTR, to compensate for low mass transfer rates in the biphasic system. The green metrics of the reaction have been assessed and compared to existing procedures and have shown the continuous process is improved over previous procedures. The organic solutions of N,N-dialkyl-N-chloramines produced continuously will enable their use in tandem flow reactions with a range of nucleophilic substrates.
International Nuclear Information System (INIS)
Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.
2010-01-01
Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.
A personal view on homogenization
International Nuclear Information System (INIS)
Tartar, L.
1987-02-01
The evolution of some ideas is first described. Under the name homogenization are collected all the mathematical results who help understanding the relations between the microstructure of a material and its macroscopic properties. Homogenization results are given through a critically detailed bibliography. The mathematical models given are systems of partial differential equations, supposed to describe some properties at a scale ε and we want to understand what will happen to the solutions if ε tends to 0
Some technical solutions on organization and technology of reactor room component mounting
International Nuclear Information System (INIS)
Romanovskij, V.I.
1982-01-01
Design of the production equipment for mounting sites of heat facilities of the Zaporozhe NPP is considered. Plan of the production equipment for mounting sites of heat facilities and flowsheet of mounting of supporting truss of the reactor are presented
Two-reactor solution spells kiss of death for the SGHWR option
International Nuclear Information System (INIS)
Butler, P.
1977-01-01
The main points from the National Nuclear Corporation's assessment of thermal reactors are summarised. Three systems SGHWR, AGR, and PWR were examined. The NNC suggests extending the AGR programme whilst at the same time developing a PWR. (U.K.)
Space reactor/organic Rankine conversion - A near-term state-of-the-art solution
Niggemann, R. E.; Lacey, D.
The use of demonstrated reactor technology with organic Rankine cycle (ORC) power conversion can provide a low cost, minimal risk approach to reactor-powered electrical generation systems in the near term. Several reactor technologies, including zirconium hydride, EBR-II and LMFBR, have demonstrated long life and suitability for space application at the operating temperature required by an efficient ORC engine. While this approach would not replace the high temperature space reactor systems presently under development, it could be available in a nearer time frame at a low and predictable cost, allowing some missions requiring high power levels to be flown prior to the availability of advanced systems with lower specific mass. Although this system has relatively high efficiency, the heat rejection temperature is low, requiring a large radiator on the order of 3.4 sq m/kWe. Therefore, a deployable heat pipe radiator configuration will be required.
DEFF Research Database (Denmark)
Shah, Vivek; Vaz Salles, Marcos António
2018-01-01
The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...
International Nuclear Information System (INIS)
Mikheev, Vladimir B.; Laulainen, Nels S.; Barlow, Stephan E.; Knott, Michael; Ford, Ian J.
2000-01-01
A laminar flow tube reactor was designed and constructed to provide an accurate, quantitative measurement of a nucleation rate as a function of supersaturation and temperature. Measurements of nucleation of a supersaturated vapor of dibutylphthalate have been made for the temperature range from -30.3 to +19.1 degree sign C. A thorough analysis of the possible sources of experimental uncertainties (such as defining the correct value of the initial vapor concentration, temperature boundary conditions on the reactor walls, accuracy of the calculations of the thermodynamic parameters of the nucleation zone, and particle concentration measurement) is given. Both isothermal and the isobaric nucleation rates were measured. The experimental data obtained were compared with the measurements of other experimental groups and with theoretical predictions made on the basis of the self-consistency correction nucleation theory. Theoretical analysis, based on the first and the second nucleation theorems, is also presented. The critical cluster size and the excess of internal energy of the critical cluster are obtained. (c) 2000 American Institute of Physics
International Nuclear Information System (INIS)
Alekseev, P.N.; Dudnikov, A.A.; Ignatiev, V.V.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.
2000-01-01
A concept of nuclear power technology system with homogeneous molten salt reactors for burning and transmutation of long-lived radioactive toxic nuclides is considered in the paper. Disposition of such reactors in enterprises of fuel cycle allows to provide them with power and facilitate solution of problems with rad waste with minimal losses. (Authors)
Annual progress report FY 1977. [Computer calculations of light water reactor dynamics and safety
Energy Technology Data Exchange (ETDEWEB)
Hansen, K.F.; Henry, A.F.
1977-07-01
Progress is summarized in a project directed toward development of numerical methods suitable for the computer solution of problems in reactor dynamics and safety. Specific areas of research include methods of integration of the time-dependent diffusion equations by finite difference and finite element methods; representation of reactor properties by various homogenization procedures; application of synthesis methods; and development of response matrix techniques.
International Nuclear Information System (INIS)
Gross, V.N.
1979-01-01
The β-activity of marked particles from the radio-chemical industry or nuclear power plants is measured in two isolated, opposed flows of homogeneous integrating mixtures. The measuring vessel for this is represented by a glass cylinder, which is separated by a glass separating wall into two parts of equal volume. The volume of the measuring vessel and therefore the volume of mixture to be measured can be increased without worsening the chromatographic separation of substances. (DG) 891 HP/DG 892 CKA [de
Solutions for Foaming Problems in Biogas Reactors Using Natural Oils or Fatty Acids as Defoamers
DEFF Research Database (Denmark)
Kougias, Panagiotis; Boe, Kanokwan; Angelidaki, Irini
2015-01-01
Foaming is one of the most common and important problems in biogas plants, leading to severe operational, economical, and environmental drawbacks. Because addition of easily degradable co-substrates for boosting the biogas production can suddenly raise the foaming problem, the full-scale biogas...... results from our previous extensive research along with some unpublished data on defoaming by rapeseed oil and oleic acid in manure-based biogas reactors. It was found that both compounds exhibited remarkable defoaming efficiency ranging from 30 to 57% in biogas reactors suffering from foaming problems...... promoted by the addition of protein, lipid, or carbohydrate co-substrates. However, in most cases, the defoaming efficiency of rapeseed oil was greater than that of oleic acid, and therefore, rapeseed oil is recommended to be used in biogas reactors to solve foaming problems....
Symmetries applied to reactor calculations
International Nuclear Information System (INIS)
Makai, M.
1982-03-01
Three problems of a reactor-calculational model are discussed with the help of symmetry considerations. 1/ A coarse mesh method applicable to any geometry is derived. It is shown that the coarse mesh solution can be constructed from a few standard boundary value problems. 2/ A second stage homogenization method is given based on the Bloch theorem. This ensures the continuity of the current and the flux at the boundary. 3/ The validity of the micro-macro separation is shown for heterogeneous lattices. A formula for the neutron density is derived for cell homogenization. (author)
International Nuclear Information System (INIS)
Silva A, L.; Del Valle G, E.
2012-10-01
This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S 2 approximation. (Author)
International Nuclear Information System (INIS)
Chapman, D.
2009-01-01
This presentation describes a dual purpose research facility at the University of Saskatchewan for Canada for the production of medical isotopes and neutrons for scientific research. The proposed research reactor is intended to supply most of Canada's medical isotope requirements and provide a neutron source for Canada's research community. Scientific research would include materials research, biomedical research and imaging.
Iterative solution to the optimal control of depletion problem in pressurized water reactors
International Nuclear Information System (INIS)
Colletti, J.P.
1981-01-01
A method is described for determining the optimal time and spatial dependence of control absorbers in the core of a pressurized water reactor over a single refueling cycle. The reactor is modeled in two dimensions with many regions using two-group diffusion theory. The problem is formulated as an optimal control problem with the cycle length fixed and the initial reactor state known. Constraints are placed on the regionwise normalized powers, control absorber concentrations, and the critical soluble boron concentration of the core. The cost functional contains two terms which may be used individually or together. One term maximizes the end-of-cycle (EOC) critical soluble boron concentration, and the other minimizes the norm of the distance between the actual and a target EOC burnup distribution. Results are given for several test problems which are based on a three-region model of the Three Mile Island Unit 1 reactor. The resulting optimal control strategies are bang-bang and lead to EOC states with the power peaking at its maximum and no control absorbers remaining in the core. Throughout the cycle the core soluble boron concentration is zero
International Nuclear Information System (INIS)
Fujibayashi, Toru.
1976-01-01
Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)
Deshwal, Bal Raj; Jin, Dong Seop; Lee, Si Hyun; Moon, Seung Hyun; Jung, Jong Hyeon; Lee, Hyung Keun
2008-02-11
The present study attempts to clean up nitric oxide from the simulated flue gas using aqueous chlorine-dioxide solution in the bubbling reactor. Chlorine-dioxide is generated by chloride-chlorate process. Experiments are carried out to examine the effect of various operating variables like input NO concentration, presence of SO(2), pH of the solution and NaCl feeding rate on the NO(x) removal efficiency at 45 degrees C. Complete oxidation of nitric oxide into nitrogen dioxide occurred on passing sufficient ClO(2) gas into the scrubbing solution. NO is finally converted into nitrate and ClO(2) is reduced into chloride ions. A plausible reaction mechanism concerning NO(x) removal by ClO(2) is suggested. DeNO(x) efficiency increased slightly with the increasing input NO concentration. The presence of SO(2) improved the NO(2) absorption but pH of solution showed marginal effect on NO(2) absorption. NO(x) removal mechanism changed when medium of solution changed from acidic to alkaline. A constant NO(x) removal efficiency of about 60% has been achieved in the wide pH range of 3-11 under optimized conditions.
Homogenized parameters of light water fuel elements computed by a perturbative (perturbation) method
International Nuclear Information System (INIS)
Koide, Maria da Conceicao Michiyo
2000-01-01
A new analytic formulation for material parameters homogenization of the two dimensional and two energy-groups diffusion model has been successfully used as a fast computational tool for recovering the detailed group fluxes in full reactor cores. The homogenization method which has been proposed does not require the solution of the diffusion problem by a numerical method. As it is generally recognized that currents at assembly boundaries must be computed accurately, a simple numerical procedure designed to improve the values of currents obtained by nodal calculations is also presented. (author)
International Nuclear Information System (INIS)
Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing
2016-01-01
Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.
Spontaneous compactification to homogeneous spaces
International Nuclear Information System (INIS)
Mourao, J.M.
1988-01-01
The spontaneous compactification of extra dimensions to compact homogeneous spaces is studied. The methods developed within the framework of coset space dimensional reduction scheme and the most general form of invariant metrics are used to find solutions of spontaneous compactification equations
A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration
International Nuclear Information System (INIS)
Dubourg, M.
1993-01-01
This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable
Energy Technology Data Exchange (ETDEWEB)
Reyes-Cruz, V.; Gonzalez, I.; Oropeza, M.T
2004-10-01
The selective electro-recovery of gold and silver values from cyanide leaching solutions containing copper was accomplished in a three-dimensional (3D) electrochemical reactor. This case let to contrast three different points of view when dealing with a composed metallic solution: First, the thermodynamic predictions; second, the microelectrolysis approach and finally, the macroelectrolysis experiments. Standard electrode potentials for the study solution would indicate a tendency for gold to deposit first. However, microelectrolysis studies of the three-metallic solution indicated that gold and silver are co-deposited onto a Vitreous carbon (VC) electrode without copper interference in a narrow potential range. Mass balances during the macroelectrolysis experiments (batch model assuming mass transfer control) indicated a preferential deposition of silver during the first ten minutes, even if gold deposition also occurred. On the other hand, values of Stanton (St) for different linear flow velocity corroborated that metals concentration gradients may establish a limit to make profitable the fluid velocity increase in an electrochemical flow cell. Electrolysis experiments were carried out under potentiostatic (at -1400 mV versus SCE) and galvanostatic (at -3.9 Am{sup -2}) conditions in the FM-01 LC flow cell.
International Nuclear Information System (INIS)
Reyes-Cruz, V.; Gonzalez, I.; Oropeza, M.T.
2004-01-01
The selective electro-recovery of gold and silver values from cyanide leaching solutions containing copper was accomplished in a three-dimensional (3D) electrochemical reactor. This case let to contrast three different points of view when dealing with a composed metallic solution: First, the thermodynamic predictions; second, the microelectrolysis approach and finally, the macroelectrolysis experiments. Standard electrode potentials for the study solution would indicate a tendency for gold to deposit first. However, microelectrolysis studies of the three-metallic solution indicated that gold and silver are co-deposited onto a Vitreous carbon (VC) electrode without copper interference in a narrow potential range. Mass balances during the macroelectrolysis experiments (batch model assuming mass transfer control) indicated a preferential deposition of silver during the first ten minutes, even if gold deposition also occurred. On the other hand, values of Stanton (St) for different linear flow velocity corroborated that metals concentration gradients may establish a limit to make profitable the fluid velocity increase in an electrochemical flow cell. Electrolysis experiments were carried out under potentiostatic (at -1400 mV versus SCE) and galvanostatic (at -3.9 Am -2 ) conditions in the FM-01 LC flow cell
International Nuclear Information System (INIS)
Marchi, Daniel E.; Menghini, Jorge E.; Trimarco, Viviana G.
1999-01-01
The inverse co-precipitation method has been used at the laboratory level to produce uranium - gadolinium mixed oxides. The formation of a mixed phase in the precipitates has been determined as well as the occurrence of only one phase in the sintered pellets, corresponding to a gadolinium - uranium solution. Moreover, a modification in the calcination-reduction stage was introduced that allows the elimination of the fissures previously detected in the sintered pellets
Solution of safety problems for nuclear power plants with WWER-440 reactors
International Nuclear Information System (INIS)
Krett, V.; Pernitsa, R.; Pfann, Ya.; Zbeglik, J.
1982-01-01
Institute of nuclear research (INR) of Czechoslovakian Atomic Energy Commission isto fulfil the supervision functions within the field of nuclear power research and development. The problems of safe operation ensurance for the nuclear power plants (NPP) with WWER-440 reactors are studied within the frame of sever major issues: code standardization and devolopment of guiding materials for the state supervision; neutronic and thermohydrolic data processing for the accident analysis; operation reliability studies of the safety systems and estimates of separate component failure importance; assessment of the accidents resulting from the equipment misfunctioning and component failures; development of a controlled reliability program; evaluation of the atomic installationsimpact on the environment; ensurence of the reactor vessel reliability and durability under irradiation. The NPP safety analysis incorporates the calculations of transient and accidental regimes for the core, the primary loop and the entire plant. A number of codes has been produced which allow to determine the state of fuel elements during operation just before the accidents assessed, thermohydrolic conditions in the coolant and the temperature distribution within the fuel both for the stationary reactor conditions and for transient regimes. A mathematical model has been deveoloped, including the description of all the primary loop major components. The Soviet code DYNAMIKA has been adopted and adjusted for EC-1040 computer, there by the accident analysis for the entire NPP has been made possible. On the basis of american SAFTE code a faster SAFEDO-2 code has been developed employing the Monte Carlo method for the accident analysis of a complex system described by means of a failure tree. The discussed codes are used at the data assessment for the accident analysis part of the safety reports as well as for the reliability evaluation of the emergency core cooling system [ru
DEFF Research Database (Denmark)
Korenak, J.; Ploder, J.; Trček, J.
2018-01-01
RNA gene and ITS1-5.8S rDNA-ITS2 sequence analysis, respectively. Serratia marcescens and Klebsiella oxytoca were the most common bacteria with the highest number present during the aerobic and anaerobic phases of the bioprocess. In addition, a high number of Elizabethkingia miricola, Morganella morganii......, Comamonas testosteroni, Trichosporon sp. and Galactomyces sp. were detected. Taken together, our results demonstrated that the sequencing batch reactor system combined with ultrafiltration is an efficient technique for treatment of wastewater containing azo dye. Moreover, the ultrafiltration effectively...
Development of Tubular Type Underwater Discharge Reactor to decompose Fe-EDTA from aqueous solution
International Nuclear Information System (INIS)
Kang, Duck-Won; Kim, Seok-Tae; Kim, Jin Kil; Ki, Hyung-Dong
2007-01-01
In case of a nuclear industry, the wastewater is hardly generated in normal operating conditions aside from laundry rooms, particularly for wastewater contaminated by radioactive materials. However if the steam generator (SG) chemical cleaning works are carrying out, it is another story. In this case we have to predict wastewater production at least from several tons to several hundreds tons during the works. Actually Kori Unit 4 in Korea is preparing the advanced sludge conditioning agents (ASCAs) project at the next overhaul period, June-2007, to remove the tube sheet scale, and we are predicting that the 200 . 250 tons waste solutions are going to produce during this works. SG chemical cleaning waste solution containing chelating agents such as EDTA is hardly easy to purify and radioactive materials included in this solution make much harder. Therefore we must have technologies to purify this chemical cleaning waste solution. The best wastewater treatment system should have great adaptability, low environmental impact, low amount of hazardous waste, and low capital and operating costs. In this study we developed the underwater spark discharge system (USDS) to decompose Fe-EDTA from aqueous solution which is contaminated with radioactive materials
International Nuclear Information System (INIS)
Park, Byungjoon; Hwang, Geelsu; Haam, Seungjoo; Lee, Changha; Ahn, Ik-Sung; Lee, Kyoungjoo
2008-01-01
Biofiltration shows high efficiency for the removal of industrial waste gases and reliable operational stability at low investment and operating cost, especially when the VOC concentration is low, such as 100 ppmv (μL L -1 ) or less. However, it has been reported that the abrupt change in VOC concentrations leads to the failure of the biofilter. Hence, the pretreatment of waste gases is necessary to ensure the stable operation of the biofilter. The objective of this study is to develop a jet loop reactor (JLR) with circulation of a surfactant solution to lower the concentration of VOCs, especially hydrophobic VOCs. Toluene and Tween 81 were used as a model industrial waste gas and a surfactant, respectively. Among several non-ionic surfactants tested, Tween 81 showed the most rapid dissolution of toluene. When a JLR is replaced with fresh Tween 81 solution (0.3% w/v) every hour, it successfully absorbed for 48 h over 90% of the toluene in an inlet gas containing toluene at 1000 ppmv (μL L -1 ) or less. Therefore, JLR with circulation of a surfactant solution is believed to ensure the stable operation of the biofilter even with the unexpected increase in the VOC concentrations
Multilevel Monte Carlo Approaches for Numerical Homogenization
Efendiev, Yalchin R.; Kronsbein, Cornelia; Legoll, Fré dé ric
2015-01-01
it comes to homogenized solutions, different levels of coarse-grid meshes are used to solve the homogenized equation. We show that, by carefully selecting the number of realizations at each level, we can achieve a speed-up in the computations in comparison
International Nuclear Information System (INIS)
Fernandez, A.; McGinley, J.; Somers, J.
2008-01-01
The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)
Energy Technology Data Exchange (ETDEWEB)
Fernandez, A.; McGinley, J.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O.Box 2340, Karlsruhe, D-76125 (Germany)
2008-07-01
The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)
International Nuclear Information System (INIS)
Chen, G.S.; Christenson, J.M.
1985-01-01
In this paper, the authors present some initial results from an investigation of the application of a locally one-dimensional (LOD) finite difference method to the solution of the two-dimensional, two-group reactor kinetics equations. Although the LOD method is relatively well known, it apparently has not been previously applied to the space-time kinetics equations. In this investigation, the LOD results were benchmarked against similar computational results (using the same computing environment, the same programming structure, and the same sample problems) obtained by the TWIGL program. For all of the problems considered, the LOD method provided accurate results in one-half to one-eight of the time required by the TWIGL program
International Nuclear Information System (INIS)
Fuchs, M.
1998-01-01
The availability depends from a lot of factors, especially from design approaches. The way for the development of design approaches for attaining the safety requirements with more redundancy and without diversity of the systems, especially of safety systems, in NPPs in the past is not acceptable for the future because of cost reasons. It is necessary to find new technical ways to achieve a better availability, reliability and safety by lower costs. The simplification of safety systems by using passive systems may be a solution in the future. The development of a new boiling water reactor BWR 1000 in Germany by Siemens which is sponsored by German utilities and supported by various European partners can be named as such example. (author)
International Nuclear Information System (INIS)
Moura Neto, C. de; Nair, R.P.K.
1979-08-01
The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt
On solution to the problem of reactor kinetics with delayed neutrons by Monte Carlo method
International Nuclear Information System (INIS)
Kyncl, Jan
2013-07-01
The initial value problem is addressed for the neutron transport equation and for the system of equations that describe the behaviour of emitters of delayed neutrons. Examination of the solution to this problem is based on several main assumptions concerning the behaviour of macroscopic effective cross-sections describing the reaction of the neutron with the medium, the temperature of medium and the remaining parameters of the equations. Formulation of these assumptions is adequately general and is in agreement with the properties of all known models of the physical quantities involved. Among others, the assumptions admit dependence of the macroscopic effective cross-sections and temperature on spatial coordinates and time that can be arbitrary to a great extent. The problem starts from a set of integro-differential equations. This problem is first transposed into the equivalent problem of solving a linear integral equation for neutron flux. This integral equation is solved by the method of successive iterations and its uniqueness is demonstrated. Numeric solution to the integral equation by Monte Carlo method consists in finding a functional of the exact solution. For this, a random process is set up and some random variables are proposed. Then it is demonstrated that each of these variables is an unbiased estimator of that functional. (author)
Olivares, Astrid; Laskin, Julia; Johnson, Grant E
2014-09-18
The scalable synthesis of ligated subnanometer metal clusters containing an exact number of atoms is of interest due to the highly size-dependent catalytic, electronic, and optical properties of these species. While significant research has been conducted on the batch preparation of clusters through reduction synthesis in solution, the processes of metal complex reduction as well as cluster nucleation, growth, and postreduction etching are still not well understood. Herein, we demonstrate a prototype temperature-controlled flow reactor for qualitatively studying cluster formation in solution at steady-state conditions. Employing this technique, methanol solutions of a chloro(triphenylphosphine)gold precursor, 1,4-bis(diphenylphosphino)butane capping ligand, and borane-tert-butylamine reducing agent were combined in a mixing tee and introduced into a heated capillary with a known length. In this manner, the temperature dependence of the relative abundance of different ionic reactants, intermediates, and products synthesized in real time was characterized qualitatively using online mass spectrometry. A wide distribution of doubly and triply charged cationic gold clusters was observed as well as smaller singly charged organometallic complexes. The results demonstrate that temperature plays a crucial role in determining the relative population of cationic gold clusters and, in general, that higher temperature promotes the formation of doubly charged clusters and singly charged organometallic complexes while reducing the abundance of triply charged species. Moreover, the distribution of clusters observed at elevated temperatures is found to be consistent with that obtained at longer reaction times at room temperature, thereby demonstrating that heating may be used to access cluster distributions characteristic of different stages of batch reduction synthesis in solution.
International Nuclear Information System (INIS)
Okuno, Hiroshi; Fujine, Yukio; Asakura, Toshihide; Murazaki, Minoru; Koyama, Tomozo; Sakakibara, Tetsuro; Shibata, Atsuhiro
1999-03-01
The crystallization method is proposed to apply for recovery of uranium from dissolution liquid, enabling to reduce handling materials in later stages of reprocessing used fast breeder reactor (FBR) fuels. This report studies possible safety problems accompanied by the proposed method. Crystallization process was first defined in the whole reprocessing process, and the quantity and the kind of treated fuel were specified. Possible problems, such as criticality, shielding, fire/explosion, and confinement, were then investigated; and the events that might induce accidental incidents were discussed. Criticality, above all the incidents, was further studied by considering exampled criticality control of the crystallization process. For crystallization equipment, in particular, evaluation models were set up in normal and accidental operation conditions. Related data were selected out from the nuclear criticality safety handbooks. The theoretical densities of plutonium nitrates, which give basic and important information, were estimated in this report based on the crystal structure data. The criticality limit of crystallization equipment was calculated based on the above information. (author)
International Nuclear Information System (INIS)
Meslin-Chiffon, E.
2007-11-01
The embrittlement of reactor pressure vessel (RPV) under irradiation is partly due to the formation of point defects (PD) and solute clusters. The aim of this work was to gain more insight into the formation mechanisms of solute clusters in low copper ([Cu] = 0.1 wt%) FeCu and FeCuMnNi model alloys, in a copper free FeMnNi model alloy and in a low copper French RPV steel (16MND5). These materials were neutron-irradiated around 300 C in a test reactor. Solute clusters were characterized by tomographic atom probe whereas PD clusters were simulated with a rate theory numerical code calibrated under cascade damage conditions using transmission electron microscopy analysis. The confrontation between experiments and simulation reveals that a heterogeneous irradiation-induced solute precipitation/segregation probably occurs on PD clusters. (author)
Formation of 32P-labelled Polyphosphates in Reactor-irradiated Solutions of Orthophosphate
DEFF Research Database (Denmark)
Fenger, Jørgen Folkvard; Pagsberg, Palle Bjørn
1973-01-01
yield increases with the concentration of the irradiated solution and varies in a complicated way with the pH. These observations and some experiments with addition of radical scavengers indicate that oxidation of the 32P-recoils by OH-radicals is an important step in the polymerization. It is suggested...... that the actual formation of a P&z.sbnd;O&z.sbnd;P bridge takes place as an addition of a Lewis acid to a lone pair of electrons on a phosphate ion....
FN approximation of the solution to a singular integral equation of classical reactor physics
Energy Technology Data Exchange (ETDEWEB)
Ganapol, B.D. [Department of Aerospace and Mechanical Engineering, University of Arizona, AME Building, Tucson, AZ 85721 (United States)]. E-mail: ganapol@ame.arizona.edu
2004-11-01
The iterated FN method is applied to a singular integral equation arising from a classical problem of reactor physics to determine the distribution of fissile material giving a spatially uniform flux. The FN iterations are accelerated toward convergence through the Wynn-algorithm - but first - Happy Birthday 'Fast Eddie' Larsen Why do I refer to the well known, most proper and exquisitely accomplished Edward W. Larsen as 'Fast Eddie'. Well our story begins in a small back bar room in the lobby of one of Los Alamos' finest and most luxurious hotels. Two young men were having a transport theoretic discussion while they were engaged in a serious game of pool with monetary benefits going to the winner. In addition, the two were sipping their most favorite lavation in rather large quantities - one, a short stocky man with thinning hair, was sipping to forget the cost of his recent divorce, and the other, a shorter stockier man also with thinning hair, was drinking, well because he liked to drink and it just made him silly. As they continued their transport discussion, one stocky man turned to the other and said, 'I wonder what 'Fast Eddie' Larsen would say to our transport question'. The other stocky man immediately thought the 'Fast Eddie' reference was to Paul Newman who played 'Fast Eddie', an expert at applied particle transport theory (a pool player) in the movie the Hustler and asked if indeed this was the case. The first stocky man said 'No. I call everyone with the name Ed 'Fast Eddie' ' - and that's the story of how 'Fast Eddie' Larsen got his name. Happy 60th Ed and thanks for all the great transport theory - from one of your biggest fans.
FN approximation of the solution to a singular integral equation of classical reactor physics
International Nuclear Information System (INIS)
Ganapol, B.D.
2004-01-01
The iterated FN method is applied to a singular integral equation arising from a classical problem of reactor physics to determine the distribution of fissile material giving a spatially uniform flux. The FN iterations are accelerated toward convergence through the Wynn-algorithm - but first - Happy Birthday 'Fast Eddie' Larsen Why do I refer to the well known, most proper and exquisitely accomplished Edward W. Larsen as 'Fast Eddie'. Well our story begins in a small back bar room in the lobby of one of Los Alamos' finest and most luxurious hotels. Two young men were having a transport theoretic discussion while they were engaged in a serious game of pool with monetary benefits going to the winner. In addition, the two were sipping their most favorite lavation in rather large quantities - one, a short stocky man with thinning hair, was sipping to forget the cost of his recent divorce, and the other, a shorter stockier man also with thinning hair, was drinking, well because he liked to drink and it just made him silly. As they continued their transport discussion, one stocky man turned to the other and said, 'I wonder what 'Fast Eddie' Larsen would say to our transport question'. The other stocky man immediately thought the 'Fast Eddie' reference was to Paul Newman who played 'Fast Eddie', an expert at applied particle transport theory (a pool player) in the movie the Hustler and asked if indeed this was the case. The first stocky man said 'No. I call everyone with the name Ed 'Fast Eddie' ' - and that's the story of how 'Fast Eddie' Larsen got his name. Happy 60th Ed and thanks for all the great transport theory - from one of your biggest fans
Multilevel Monte Carlo Approaches for Numerical Homogenization
Efendiev, Yalchin R.
2015-10-01
In this article, we study the application of multilevel Monte Carlo (MLMC) approaches to numerical random homogenization. Our objective is to compute the expectation of some functionals of the homogenized coefficients, or of the homogenized solutions. This is accomplished within MLMC by considering different sizes of representative volumes (RVEs). Many inexpensive computations with the smallest RVE size are combined with fewer expensive computations performed on larger RVEs. Likewise, when it comes to homogenized solutions, different levels of coarse-grid meshes are used to solve the homogenized equation. We show that, by carefully selecting the number of realizations at each level, we can achieve a speed-up in the computations in comparison to a standard Monte Carlo method. Numerical results are presented for both one-dimensional and two-dimensional test-cases that illustrate the efficiency of the approach.
International Nuclear Information System (INIS)
Venter, A.M.
1973-08-01
A short discussion is given of the physics of a nuclear reactor and the parameters which are used in the study of neutron transport. The mathematical formulation and detailed derivation is given of the neutron diffusion and transport equations. A description is given of the computer programmes, FIRE-5 and PELSN, developed at Pelindaba for the evaluation of both thermal and fast reactor systems. It is indicated how these computer programmes have been applied in the study of the PELINDUNA-O and other known critical facilities. The application of Lie-series to the solution of the neutron diffusion equation is discussed in detail. The time dependence of the variables is removed by means of a Laplacetransformation and the semi-analytical solution is written in terms of a transfer matrix. A complete set of recursion formulae, applicable to both homogeneous and heterogeneous reactor systems, is derived. The method used in the evaluation of the effective multiplication factor, k-eff, and the alpha-eigen-value is described. A computer programme was written to solve the neutron diffusion equation in terms of the Lie-series. The results are compared with the TIMOC and PELSN computer programmes. A method is suggested in which the Lie-series are used to solve the neutron transport equation. The transfer matrix for this case, is derived. A complete discussion is given of the solution to the space and time dependent diffusion equation in the presence of a delta source [af
International Nuclear Information System (INIS)
Kasper, K.J.
1975-06-01
In the field of reactor engineering an increasing tendency is visible towards a 'repairable reactor'. In the construction of the HTR with spherical fuel elements this fact should already be taken into account at an early stage. Additionally it is possible that in connection with the OTTO-fueling load conditions for the graphite reflector could result which are locally not far away from limiting values. Therefore the removability of the reflector is included in the reactor construction as an accompanying technical step of the physical lay-out of the core. The core arrangements, realized for HTR until recently, are discussed as well as the properties of the graphites used and the operating conditions in the reactors are stated. At the example of the PR 3,000 proposals are offered for the construction of a removable side and top reflector for a pebble bed reactor. Hereby a solution was found which, on one hand allows the changing of the reflector and on the other hand requires no significant increase of the costs for the reactor assembly. Moreover the requirements of reactor operation and of repairability are satisfied in an optimal manner. (orig.) [de
Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics
International Nuclear Information System (INIS)
Henry, A.F.
1980-01-01
Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented
Mechanical Homogenization Increases Bacterial Homogeneity in Sputum
Stokell, Joshua R.; Khan, Ammad
2014-01-01
Sputum obtained from patients with cystic fibrosis (CF) is highly viscous and often heterogeneous in bacterial distribution. Adding dithiothreitol (DTT) is the standard method for liquefaction prior to processing sputum for molecular detection assays. To determine if DTT treatment homogenizes the bacterial distribution within sputum, we measured the difference in mean total bacterial abundance and abundance of Burkholderia multivorans between aliquots of DTT-treated sputum samples with and without a mechanical homogenization (MH) step using a high-speed dispersing element. Additionally, we measured the effect of MH on bacterial abundance. We found a significant difference between the mean bacterial abundances in aliquots that were subjected to only DTT treatment and those of the aliquots which included an MH step (all bacteria, P = 0.04; B. multivorans, P = 0.05). There was no significant effect of MH on bacterial abundance in sputum. Although our results are from a single CF patient, they indicate that mechanical homogenization increases the homogeneity of bacteria in sputum. PMID:24759710
Energy Technology Data Exchange (ETDEWEB)
Capdebosq, Y
1999-09-01
In order to study and simulate nuclear reactor cores, one needs to access the neutron distribution in the core. In practice, the description of this density of neutrons is given by a system of diffusion equations, coupled by non differential exchange terms. The strong heterogeneity of the medium constitutes a major obstacle to the numerical computation of this models at reasonable cost. Homogenization appears as compulsory. Heuristic methods have been developed since the origin by nuclear physicists, under a periodicity assumption on the coefficients. They consist in doing a fine computation one a single periodicity cell, to solve the system on the whole domain with homogeneous coefficients, and to reconstruct the neutron density by multiplying the solutions of the two computations. The objectives of this work are to provide mathematically rigorous basis to this factorization method, to obtain the exact formulas of the homogenized coefficients, and to start on geometries where two periodical medium are placed side by side. The first result of this thesis concerns eigenvalue problem models which are used to characterize the state of criticality of the reactor, under a symmetry assumption on the coefficients. The convergence of the homogenization process is proved, and formulas of the homogenized coefficients are given. We then show that without symmetry assumptions, a drift phenomenon appears. It is characterized by the mean of a real Bloch wave method, which gives the homogenized limit in the general case. These results for the critical problem are then adapted to the evolution model. Finally, the homogenization of the critical problem in the case of two side by side periodic medium is studied on a one dimensional on equation model. (authors)
Functionality and homogeneity.
2011-01-01
Functionality and homogeneity are two of the five Sustainable Safety principles. The functionality principle aims for roads to have but one exclusive function and distinguishes between traffic function (flow) and access function (residence). The homogeneity principle aims at differences in mass,
Rapid hydrolysis of celluloses in homogeneous solution
Energy Technology Data Exchange (ETDEWEB)
Garves, K
1979-01-01
Dissolution of cellulose (I), cotton, and cotton linters in a mixture of Ac0H, Ac/sub 2/O, H/sub 2/SO/sub 4/, and DMF at 120 to 160 degrees resulted in rapid and complete hydrolysis of I with decomposition of the cellulose acetatesulfate formed by gradual addition of aqueous acid. Highly crystalline I is quickly decomposed to glucose with minimum byproduct formation. Carbohydrate products containing sugar units other than glucose are hydrolyzed with destruction of monosaccharides.
Homogenization of Mammalian Cells.
de Araújo, Mariana E G; Lamberti, Giorgia; Huber, Lukas A
2015-11-02
Homogenization is the name given to the methodological steps necessary for releasing organelles and other cellular constituents as a free suspension of intact individual components. Most homogenization procedures used for mammalian cells (e.g., cavitation pump and Dounce homogenizer) rely on mechanical force to break the plasma membrane and may be supplemented with osmotic or temperature alterations to facilitate membrane disruption. In this protocol, we describe a syringe-based homogenization method that does not require specialized equipment, is easy to handle, and gives reproducible results. The method may be adapted for cells that require hypotonic shock before homogenization. We routinely use it as part of our workflow to isolate endocytic organelles from mammalian cells. © 2015 Cold Spring Harbor Laboratory Press.
Yamanaka, Ichiro; Onisawa, Takeshi; Hashimoto, Toshikazu; Murayama, Toru
2011-04-18
The effects of the type of fuel-cell reactors (undivided or divided by cation- and anion-exchange membranes), alkaline electrolytes (LiOH, NaOH, KOH), vapor-grown carbon fiber (VGCF) cathode components (additives: none, activated carbon, Valcan XC72, Black Pearls 2000, Seast-6, and Ketjen Black), and the flow rates of anolyte (0, 1.5, 12 mL h(-1)) and catholyte (0, 12 mL h(-1)) on the formation of hydrogen peroxide were studied. A divided fuel-cell system, O(2) (g)|VGCF-XC72 cathode|2 M NaOH catholyte|cation-exchange membrane (Nafion-117)|Pt/XC72-VGCF anode|2 M NaOH anolyte at 12 mL h(-1) flow|H(2) (g), was effective for the selective formation of hydrogen peroxide, with 130 mA cm(-2) , a 2 M aqueous solution of H(2)O(2)/NaOH, and a current efficiency of 95 % at atmospheric pressure and 298 K. The current and formation rate gradually decreased over a long period of time. The cause of the slow decrease in electrocatalytic performance was revealed and the decrease was stopped by a flow of catholyte. Cyclic voltammetry studies at the VGCF-XC72 electrode indicated that fast diffusion of O(2) from the gas phase to the electrode, and quick desorption of hydrogen peroxide from the electrode to the electrolyte were essential for the efficient formation of solutions of H(2)O(2)/NaOH. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
International Nuclear Information System (INIS)
Meneses, Anderson Alvarenga de Moura; Schirru, Roberto
2005-01-01
This work focuses on the usage the Artificial Intelligence technique Particle Swarm Optimization (PSO) to optimize the fuel recharge at a nuclear reactor. This is a combinatorial problem, in which the search of the best feasible solution is done by minimizing a specific objective function. However, in this first moment it is possible to compare the fuel recharge problem with the Traveling Salesman Problem (TSP), since both of them are combinatorial, with one advantage: the evaluation of the TSP objective function is much more simple. Thus, the proposed methods have been applied to two TSPs: Oliver 30 and Rykel 48. In 1995, KENNEDY and EBERHART presented the PSO technique to optimize non-linear continued functions. Recently some PSO models for discrete search spaces have been developed for combinatorial optimization. Although all of them having different formulation from the ones presented here. In this paper, we use the PSO theory associated with to the Random Keys (RK)model, used in some optimizations with Genetic Algorithms. The Particle Swarm Optimization with Random Keys (PSORK) results from this association, which combines PSO and RK. The adaptations and changings in the PSO aim to allow the usage of the PSO at the nuclear fuel recharge. This work shows the PSORK being applied to the proposed combinatorial problem and the obtained results. (author)
Mach's principle in spatially homogeneous spacetimes
International Nuclear Information System (INIS)
Tipler, F.J.
1978-01-01
On the basis of Mach's Principle it is concluded that the only singularity-free solution to the empty space Einstein equations is flat space. It is shown that the only singularity-free solution to the empty space Einstein equations which is spatially homogeneous and globally hyperbolic is in fact suitably identified Minkowski space. (Auth.)
The SPH homogeneization method
International Nuclear Information System (INIS)
Kavenoky, Alain
1978-01-01
The homogeneization of a uniform lattice is a rather well understood topic while difficult problems arise if the lattice becomes irregular. The SPH homogeneization method is an attempt to generate homogeneized cross sections for an irregular lattice. Section 1 summarizes the treatment of an isolated cylindrical cell with an entering surface current (in one velocity theory); Section 2 is devoted to the extension of the SPH method to assembly problems. Finally Section 3 presents the generalisation to general multigroup problems. Numerical results are obtained for a PXR rod bundle assembly in Section 4
Homogeneity of Inorganic Glasses
DEFF Research Database (Denmark)
Jensen, Martin; Zhang, L.; Keding, Ralf
2011-01-01
Homogeneity of glasses is a key factor determining their physical and chemical properties and overall quality. However, quantification of the homogeneity of a variety of glasses is still a challenge for glass scientists and technologists. Here, we show a simple approach by which the homogeneity...... of different glass products can be quantified and ranked. This approach is based on determination of both the optical intensity and dimension of the striations in glasses. These two characteristic values areobtained using the image processing method established recently. The logarithmic ratio between...
Flows and chemical reactions in homogeneous mixtures
Prud'homme, Roger
2013-01-01
Flows with chemical reactions can occur in various fields such as combustion, process engineering, aeronautics, the atmospheric environment and aquatics. The examples of application chosen in this book mainly concern homogeneous reactive mixtures that can occur in propellers within the fields of process engineering and combustion: - propagation of sound and monodimensional flows in nozzles, which may include disequilibria of the internal modes of the energy of molecules; - ideal chemical reactors, stabilization of their steady operation points in the homogeneous case of a perfect mixture and c
Considerations in the design of a high power medical isotope production reactor
International Nuclear Information System (INIS)
Ball, Russell M.; Nordyke, William H.; Brown, Roy
2002-01-01
For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)
Layout optimization using the homogenization method
Suzuki, Katsuyuki; Kikuchi, Noboru
1993-01-01
A generalized layout problem involving sizing, shape, and topology optimization is solved by using the homogenization method for three-dimensional linearly elastic shell structures in order to seek a possibility of establishment of an integrated design system of automotive car bodies, as an extension of the previous work by Bendsoe and Kikuchi. A formulation of a three-dimensional homogenized shell, a solution algorithm, and several examples of computing the optimum layout are presented in this first part of the two articles.
Benchmarking homogenization algorithms for monthly data
Venema, V. K. C.; Mestre, O.; Aguilar, E.; Auer, I.; Guijarro, J. A.; Domonkos, P.; Vertacnik, G.; Szentimrey, T.; Stepanek, P.; Zahradnicek, P.; Viarre, J.; Müller-Westermeier, G.; Lakatos, M.; Williams, C. N.; Menne, M. J.; Lindau, R.; Rasol, D.; Rustemeier, E.; Kolokythas, K.; Marinova, T.; Andresen, L.; Acquaotta, F.; Fratiannil, S.; Cheval, S.; Klancar, M.; Brunetti, M.; Gruber, C.; Prohom Duran, M.; Likso, T.; Esteban, P.; Brandsma, T.; Willett, K.
2013-09-01
The COST (European Cooperation in Science and Technology) Action ES0601: Advances in homogenization methods of climate series: an integrated approach (HOME) has executed a blind intercomparison and validation study for monthly homogenization algorithms. Time series of monthly temperature and precipitation were evaluated because of their importance for climate studies. The algorithms were validated against a realistic benchmark dataset. Participants provided 25 separate homogenized contributions as part of the blind study as well as 22 additional solutions submitted after the details of the imposed inhomogeneities were revealed. These homogenized datasets were assessed by a number of performance metrics including i) the centered root mean square error relative to the true homogeneous values at various averaging scales, ii) the error in linear trend estimates and iii) traditional contingency skill scores. The metrics were computed both using the individual station series as well as the network average regional series. The performance of the contributions depends significantly on the error metric considered. Although relative homogenization algorithms typically improve the homogeneity of temperature data, only the best ones improve precipitation data. Moreover, state-of-the-art relative homogenization algorithms developed to work with an inhomogeneous reference are shown to perform best. The study showed that currently automatic algorithms can perform as well as manual ones.
International Nuclear Information System (INIS)
Schulze, I.; Gutscher, E.
1980-01-01
The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de
Application of cellular neural network (CNN) method to the nuclear reactor dynamics equations
International Nuclear Information System (INIS)
Hadad, K.; Piroozmand, A.
2007-01-01
This paper describes the application of a multilayer cellular neural network (CNN) to model and solve the nuclear reactor dynamic equations. An equivalent electrical circuit is analyzed and the governing equations of a bare, homogeneous reactor core are modeled via CNN. The validity of the CNN result is compared with numerical solution of the system of nonlinear governing partial differential equations (PDE) using MATLAB. Steady state as well as transient simulations, show very good comparison between the two methods. We used our CNN model to simulate space-time response of different reactivity excursions in a typical nuclear reactor. On line solution of reactor dynamic equations is used as an aid to reactor operation decision making. The complete algorithm could also be implemented using very large scale integrated circuit (VLSI) circuitry. The efficiency of the calculation method makes it useful for small size nuclear reactors such as the ones used in space missions
Homogenized group cross sections by Monte Carlo
International Nuclear Information System (INIS)
Van Der Marck, S. C.; Kuijper, J. C.; Oppe, J.
2006-01-01
Homogenized group cross sections play a large role in making reactor calculations efficient. Because of this significance, many codes exist that can calculate these cross sections based on certain assumptions. However, the application to the High Flux Reactor (HFR) in Petten, the Netherlands, the limitations of such codes imply that the core calculations would become less accurate when using homogenized group cross sections (HGCS). Therefore we developed a method to calculate HGCS based on a Monte Carlo program, for which we chose MCNP. The implementation involves an addition to MCNP, and a set of small executables to perform suitable averaging after the MCNP run(s) have completed. Here we briefly describe the details of the method, and we report on two tests we performed to show the accuracy of the method and its implementation. By now, this method is routinely used in preparation of the cycle to cycle core calculations for HFR. (authors)
International Nuclear Information System (INIS)
Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana
2006-01-01
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments
Diffusion piecewise homogenization via flux discontinuity ratios
International Nuclear Information System (INIS)
Sanchez, Richard; Dante, Giorgio; Zmijarevic, Igor
2013-01-01
We analyze piecewise homogenization with flux-weighted cross sections and preservation of averaged currents at the boundary of the homogenized domain. Introduction of a set of flux discontinuity ratios (FDR) that preserve reference interface currents leads to preservation of averaged region reaction rates and fluxes. We consider the class of numerical discretizations with one degree of freedom per volume and per surface and prove that when the homogenization and computing meshes are equal there is a unique solution for the FDRs which exactly preserve interface currents. For diffusion sub-meshing we introduce a Jacobian-Free Newton-Krylov method and for all cases considered obtain an 'exact' numerical solution (eight digits for the interface currents). The homogenization is completed by extending the familiar full assembly homogenization via flux discontinuity factors to the sides of regions laying on the boundary of the piecewise homogenized domain. Finally, for the familiar nodal discretization we numerically find that the FDRs obtained with no sub-mesh (nearly at no cost) can be effectively used for whole-core diffusion calculations with sub-mesh. This is not the case, however, for cell-centered finite differences. (authors)
Dynamics of homogeneous nucleation
DEFF Research Database (Denmark)
Toxværd, Søren
2015-01-01
The classical nucleation theory for homogeneous nucleation is formulated as a theory for a density fluctuation in a supersaturated gas at a given temperature. But molecular dynamics simulations reveal that it is small cold clusters which initiates the nucleation. The temperature in the nucleating...
Homogeneous bilateral block shifts
Indian Academy of Sciences (India)
Douglas class were classified in [3]; they are unilateral block shifts of arbitrary block size (i.e. dim H(n) can be anything). However, no examples of irreducible homogeneous bilateral block shifts of block size larger than 1 were known until now.
Tignanelli, H. L.; Vazquez, R. A.; Mostaccio, C.; Gordillo, S.; Plastino, A.
1990-11-01
RESUMEN. Presentamos una metodologia de analisis de la homogeneidad a partir de la Teoria de la Informaci6n, aplicable a muestras de datos observacionales. ABSTRACT:Standard concepts that underlie Information Theory are employed in order design a methodology that enables one to analyze the homogeneity of a given data sample. Key : DATA ANALYSIS
Homogeneous Poisson structures
International Nuclear Information System (INIS)
Shafei Deh Abad, A.; Malek, F.
1993-09-01
We provide an algebraic definition for Schouten product and give a decomposition for any homogenenous Poisson structure in any n-dimensional vector space. A large class of n-homogeneous Poisson structures in R k is also characterized. (author). 4 refs
International Nuclear Information System (INIS)
Nakahara, Yasuaki; Ise, Takeharu; Kobayashi, Kensuke; Itoh, Yasuyuki
1975-12-01
A new method has been developed for numerical solution of a class of nonlinear Volterra integro-differential equations with quadratic nonlinearity. After dividing the domain of the variable into subintervals, piecewise approximations are applied in the subintervals. The equation is first integrated over a subinterval to obtain the piecewise equation, to which six approximate treatments are applied, i.e. fully explicit, fully implicit, Crank-Nicolson, linear interpolation, quadratic and cubic spline. The numerical solution at each time step is obtained directly as a positive root of the resulting algebraic quadratic equation. The point reactor kinetics with a ramp reactivity insertion, linear temperature feedback and delayed neutrons can be described by one of this type of nonlinear Volterra integro-differential equations. The algorithm is applied to the Argonne benchmark problem and a model problem for a fast reactor without delayed neutrons. The fully implicit method has been found to be unconditionally stable in the sense that it always gives the positive real roots. The cubic spline method is divergent, and the other four methods are intermediate in between. From the estimation of the stability, convergency, accuracy and CPU time, it is concluded that the Crank-Nicolson method is best, then the linear interpolation method comes closely next to it. Discussions are also made on the possibility of applying the algorithm to the fusion reactor kinetics in the form of a nonlinear partial differential equation. (auth.)
Energy Technology Data Exchange (ETDEWEB)
Silva A, L.; Del Valle G, E., E-mail: evalle@ipn.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico)
2012-10-15
This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S{sub 2} approximation. (Author)
Benchmarking homogenization algorithms for monthly data
Directory of Open Access Journals (Sweden)
V. K. C. Venema
2012-01-01
Full Text Available The COST (European Cooperation in Science and Technology Action ES0601: advances in homogenization methods of climate series: an integrated approach (HOME has executed a blind intercomparison and validation study for monthly homogenization algorithms. Time series of monthly temperature and precipitation were evaluated because of their importance for climate studies and because they represent two important types of statistics (additive and multiplicative. The algorithms were validated against a realistic benchmark dataset. The benchmark contains real inhomogeneous data as well as simulated data with inserted inhomogeneities. Random independent break-type inhomogeneities with normally distributed breakpoint sizes were added to the simulated datasets. To approximate real world conditions, breaks were introduced that occur simultaneously in multiple station series within a simulated network of station data. The simulated time series also contained outliers, missing data periods and local station trends. Further, a stochastic nonlinear global (network-wide trend was added.
Participants provided 25 separate homogenized contributions as part of the blind study. After the deadline at which details of the imposed inhomogeneities were revealed, 22 additional solutions were submitted. These homogenized datasets were assessed by a number of performance metrics including (i the centered root mean square error relative to the true homogeneous value at various averaging scales, (ii the error in linear trend estimates and (iii traditional contingency skill scores. The metrics were computed both using the individual station series as well as the network average regional series. The performance of the contributions depends significantly on the error metric considered. Contingency scores by themselves are not very informative. Although relative homogenization algorithms typically improve the homogeneity of temperature data, only the best ones improve
Homogeneous group, research, institution
Directory of Open Access Journals (Sweden)
Francesca Natascia Vasta
2014-09-01
Full Text Available The work outlines the complex connection among empiric research, therapeutic programs and host institution. It is considered the current research state in Italy. Italian research field is analyzed and critic data are outlined: lack of results regarding both the therapeutic processes and the effectiveness of eating disorders group analytic treatment. The work investigates on an eating disorders homogeneous group, led into an eating disorder outpatient service. First we present the methodological steps the research is based on including the strong connection among theory and clinical tools. Secondly clinical tools are described and the results commented. Finally, our results suggest the necessity of validating some more specifical hypothesis: verifying the relationship between clinical improvement (sense of exclusion and painful emotions reduction and specific group therapeutic processes; verifying the relationship between depressive feelings, relapses and transition trough a more differentiated groupal field.Keywords: Homogeneous group; Eating disorders; Institutional field; Therapeutic outcome
Homogeneous turbulence dynamics
Sagaut, Pierre
2018-01-01
This book provides state-of-the-art results and theories in homogeneous turbulence, including anisotropy and compressibility effects with extension to quantum turbulence, magneto-hydodynamic turbulence and turbulence in non-newtonian fluids. Each chapter is devoted to a given type of interaction (strain, rotation, shear, etc.), and presents and compares experimental data, numerical results, analysis of the Reynolds stress budget equations and advanced multipoint spectral theories. The role of both linear and non-linear mechanisms is emphasized. The link between the statistical properties and the dynamics of coherent structures is also addressed. Despite its restriction to homogeneous turbulence, the book is of interest to all people working in turbulence, since the basic physical mechanisms which are present in all turbulent flows are explained. The reader will find a unified presentation of the results and a clear presentation of existing controversies. Special attention is given to bridge the results obta...
Homogen Mur - et udviklingsprojekt
DEFF Research Database (Denmark)
Dahl, Torben; Beim, Anne; Sørensen, Peter
1997-01-01
Mølletorvet i Slagelse er det første byggeri i Danmark, hvor ydervæggen er udført af homogene bærende og isolerende teglblokke. Byggeriet viser en række af de muligheder, der både med hensyn til konstruktioner, energiforhold og arkitektur ligger i anvendelsen af homogent blokmurværk.......Mølletorvet i Slagelse er det første byggeri i Danmark, hvor ydervæggen er udført af homogene bærende og isolerende teglblokke. Byggeriet viser en række af de muligheder, der både med hensyn til konstruktioner, energiforhold og arkitektur ligger i anvendelsen af homogent blokmurværk....
Homogenization of resonant chiral metamaterials
DEFF Research Database (Denmark)
Andryieuski, Andrei; Menzel, C.; Rockstuhl, Carsten
2010-01-01
Homogenization of metamaterials is a crucial issue as it allows to describe their optical response in terms of effective wave parameters as, e.g., propagation constants. In this paper we consider the possible homogenization of chiral metamaterials. We show that for meta-atoms of a certain size...... an analytical criterion for performing the homogenization and a tool to predict the homogenization limit. We show that strong coupling between meta-atoms of chiral metamaterials may prevent their homogenization at all....
International Nuclear Information System (INIS)
Figueroa-O’Farrill, José; Ungureanu, Mara
2016-01-01
Motivated by the search for new gravity duals to M2 branes with N>4 supersymmetry — equivalently, M-theory backgrounds with Killing superalgebra osp(N|4) for N>4 — we classify (except for a small gap) homogeneous M-theory backgrounds with symmetry Lie algebra so(n)⊕so(3,2) for n=5,6,7. We find that there are no new backgrounds with n=6,7 but we do find a number of new (to us) backgrounds with n=5. All backgrounds are metrically products of the form AdS 4 ×P 7 , with P riemannian and homogeneous under the action of SO(5), or S 4 ×Q 7 with Q lorentzian and homogeneous under the action of SO(3,2). At least one of the new backgrounds is supersymmetric (albeit with only N=2) and we show that it can be constructed from a supersymmetric Freund-Rubin background via a Wick rotation. Two of the new backgrounds have only been approximated numerically.
Energy Technology Data Exchange (ETDEWEB)
Figueroa-O’Farrill, José [School of Mathematics and Maxwell Institute for Mathematical Sciences,The University of Edinburgh,James Clerk Maxwell Building, The King’s Buildings, Peter Guthrie Tait Road,Edinburgh EH9 3FD, Scotland (United Kingdom); Ungureanu, Mara [Humboldt-Universität zu Berlin, Institut für Mathematik,Unter den Linden 6, 10099 Berlin (Germany)
2016-01-25
Motivated by the search for new gravity duals to M2 branes with N>4 supersymmetry — equivalently, M-theory backgrounds with Killing superalgebra osp(N|4) for N>4 — we classify (except for a small gap) homogeneous M-theory backgrounds with symmetry Lie algebra so(n)⊕so(3,2) for n=5,6,7. We find that there are no new backgrounds with n=6,7 but we do find a number of new (to us) backgrounds with n=5. All backgrounds are metrically products of the form AdS{sub 4}×P{sup 7}, with P riemannian and homogeneous under the action of SO(5), or S{sup 4}×Q{sup 7} with Q lorentzian and homogeneous under the action of SO(3,2). At least one of the new backgrounds is supersymmetric (albeit with only N=2) and we show that it can be constructed from a supersymmetric Freund-Rubin background via a Wick rotation. Two of the new backgrounds have only been approximated numerically.
International Nuclear Information System (INIS)
Conti, C.F.S.; Watson, F.V.
1991-01-01
A computational code to solve a two energy group neutron diffusion problem has been developed base d on the Response Matrix Method. That method solves the global problem of PWR core, without using the cross sections homogenization process, thus it is equivalent to a pontwise core calculation. The present version of the code calculates the response matrices by the first order perturbative method and considers developments on arbitrary order Fourier series for the boundary fluxes and interior fluxes. (author)
International Nuclear Information System (INIS)
Akaho, E.H.K.; Danso, K.A.
1990-07-01
The code ISODEP is being developed to compute the rate of production of nuclides in different homogeneous zones of a research reactor. ISODEP will form part of a main programme MEPE for analysis of the Miniature Neutron Source Reactor. An exponential method proposed by Hansen is the basis for the numerical solution of non-homogeneous simultaneous equations which describe the rate of production and decay of nuclides. Four principal chains associated with uranium-fuelled reactor were studied. The trend of results was found to be consistent with those obtained by analytical method. It is hoped that after slight modifications of the code and with appropriate effective microscopic cross-section data it will be suitable for research reactor analysis. (author)
Study on critical effect in lattice homogenization via Monte Carlo method
International Nuclear Information System (INIS)
Li Mancang; Wang Kan; Yao Dong
2012-01-01
In contrast to the traditional deterministic lattice codes, generating the homogenization multigroup constants via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum. thus provides more accuracy parameters. An infinite lattice of identical symmetric motives is usually assumed when performing the homogenization. However, the finite size of a reactor is reality and it should influence the lattice calculation. In practice of the homogenization with Monte Carlo method, B N theory is applied to take the leakage effect into account. The fundamental mode with the buckling B is used as a measure of the finite size. The critical spectrum in the solution of 0-dimensional fine-group B 1 equations is used to correct the weighted spectrum for homogenization. A PWR prototype core is examined to verify that the presented method indeed generates few group constants effectively. In addition, a zero power physical experiment verification is performed. The results show that B N theory is adequate for leakage correction in the multigroup constants generation via Monte Carlo method. (authors)
Deng, Shaoqiang
2012-01-01
"Homogeneous Finsler Spaces" is the first book to emphasize the relationship between Lie groups and Finsler geometry, and the first to show the validity in using Lie theory for the study of Finsler geometry problems. This book contains a series of new results obtained by the author and collaborators during the last decade. The topic of Finsler geometry has developed rapidly in recent years. One of the main reasons for its surge in development is its use in many scientific fields, such as general relativity, mathematical biology, and phycology (study of algae). This monograph introduc
Homogeneity spoil spectroscopy
International Nuclear Information System (INIS)
Hennig, J.; Boesch, C.; Martin, E.; Grutter, R.
1987-01-01
One of the problems of in vivo MR spectroscopy of P-31 is spectra localization. Surface coil spectroscopy, which is the method of choice for clinical applications, suffers from the high-intensity signal from subcutaneous muscle tissue, which masks the spectrum of interest from deeper structures. In order to suppress this signal while maintaining the simplicity of surface coil spectroscopy, the authors introduced a small sheet of ferromagnetically dotted plastic between the surface coil and the body. This sheet destroys locally the field homogeneity and therefore all signal from structures around the coil. The very high reproducibility of the simple experimental procedure allows long-term studies important for monitoring tumor therapy
International Nuclear Information System (INIS)
Craemer, B.; Dahm, H.; Spillekothen, H.G.
1982-06-01
The design basis of the reactor protection system (RPS) for HTR plants generating process heat and electric power is briefly described and some particularities of process heat plants are indicated. Some particularly important or exacting technical measuring positions for the RPS of a process heat HTR with 500 MWsub(th) power (PNP 500) are described and current R + D work explained. It is demonstrated that a particularly simple RPS can be realized in an HTR with modular design. (author)
International Nuclear Information System (INIS)
Suri, Rominder P.S.; Nayak, Mohan; Devaiah, Uthappa; Helmig, Edward
2007-01-01
There are many reports documenting the adverse effects, such as feminization of fish, of estrogen hormones in the environment. One of the major sources of these compounds is from municipal wastewater effluents. The biological processes at municipal wastewater treatment plants cannot completely remove these compounds. This paper discusses the use of ultrasound to destroy estrogen compounds in water. The study examines the effect of ultrasound power density and power intensity on the destruction of various estrogen compounds which include: 17α-estradiol, 17β-estradiol, estrone, estriol, equilin, 17α-dihydroequilin, 17α-ethinyl estradiol and norgestrel. These tests were conducted in single component batch and flow through reactors using 0.6, 2 and 4 kW ultrasound sources. The sonolysis process produced 80-90% destruction of individual estrogens at initial concentration of 10 μg/L within 40-60 min of contact time. First order rate constants for the individual compounds under different conditions are presented. The estrogen degradation rates increase with increase in power intensity. However, the energy efficiency of the reactor was higher at lower power density. The 4 kW ultrasound reactor was more energy efficient compared to the 0.6 and 2 kW sonicators
Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4
International Nuclear Information System (INIS)
Martinez F, M.A.; Valle G, E. del; Alonso V, G.
2007-01-01
In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)
Developing a multi-physics solver in APOLLO3 and applications to cross section homogenization
International Nuclear Information System (INIS)
Dugan, Kevin-James
2016-01-01
Multi-physics coupling is becoming of large interest in the nuclear engineering and computational science fields. The ability to obtain accurate solutions to realistic models is important to the design and licensing of novel reactor designs, especially in design basis accident situations. The physical models involved in calculating accident behavior in nuclear reactors includes: neutron transport, thermal conduction/convection, thermo-mechanics in fuel and support structure, fuel stoichiometry, among others. However, this thesis focuses on the coupling between two models, neutron transport and thermal conduction/convection.The goal of this thesis is to develop a multi-physics solver for simulating accidents in nuclear reactors. The focus is both on the simulation environment and the data treatment used in such simulations.This work discusses the development of a multi-physics framework based around the Jacobian-Free Newton-Krylov (JFNK) method. The framework includes linear and nonlinear solvers, along with interfaces to existing numerical codes that solve neutron transport and thermal hydraulics models (APOLLO3 and MCTH respectively) through the computation of residuals. a new formulation for the neutron transport residual is explored, which reduces the solution size and search space by a large factor; instead of the residual being based on the angular flux, it is based on the fission source.The question of whether using a fundamental mode distribution of the neutron flux for cross section homogenization is sufficiently accurate during fast transients is also explored. It is shown that in an infinite homogeneous medium, using homogenized cross sections produced with a fundamental mode flux differ significantly from a reference solution. The error is remedied by using an alternative weighting flux taken from a time dependent calculation; either a time-integrated flux or an asymptotic solution. The time-integrated flux comes from the multi-physics solution of the
Safeguarding research reactors
International Nuclear Information System (INIS)
Powers, J.A.
1983-03-01
The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor
International Nuclear Information System (INIS)
Carreira, M.
1965-01-01
In order to reduce limitations of solubility, the cryoscopic method developed for benzene solutions of polyphenyl mixtures has been extended to diphenyl-ether solutions by introducing some modifications imposed by the physico-chemical properties of this solvent. The Nernsto theory of Beckman's method has been revised, taking into account the heat-transfer characteristics of the system, and the results of that analysis have been used to fix upon the design parameters of a cryoscopic apparatus for measurements on diphenyl-ether solutions. (Author) 9 refs
International Nuclear Information System (INIS)
Vo Thi Cam Hoa; Duong Van Dong; Nguyen Thi Thu; Chu Van Khoa
2007-01-01
This report describes the practical methods for analyzing of Tellurium content in Na 131 I solution produced at the Dalat Nuclear Research Institute. We studied analytical methods to control Tellurium content in final Na 131 I solution product used in medical purposes by three methods such as: spot test, gamma spectrometric and spectrophotometric methods. These investigation results are shown that the spot test method is suitable for controlling Tellurium trace in the final product. This spot test can be determinate Tellurium trace less than 10 ppm and are used to quality control of Na 131 I solution using in medical application. (author)
Directory of Open Access Journals (Sweden)
Z. Gholamzadeh
2015-12-01
Full Text Available The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99Mo. In this method, the medical isotope production system itself is used to extract 99Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99Mo by irradiating targets. In this study, the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ∼1,500 Ci/wk (∼325 6-day Ci of 99Mo at the end of a cycle.
Analysis of short-term reactor cavity transient
International Nuclear Information System (INIS)
Cheng, T.C.; Fischer, S.R.
1981-01-01
Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity
MOX in reactors: present and future
International Nuclear Information System (INIS)
Arslan, Marc; Gros, Jean Pierre; Niquille, Aurelie; Marincic, Alexis
2010-01-01
In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR TM or ATMEA TM designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR TM reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR TM can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO 2 , ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)
Transmutation of Americium in Light and Heavy Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)
2009-06-15
There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)
International Nuclear Information System (INIS)
Kallenbach, A.; Bosch, H.-S.; De Pena Hempel, S.; Dux, R.; Kaufmann, M.; Mertens, V.; Neuhauser, J.; Suttrop, W.; Zohm, H.
1997-01-01
For pt.II see ibid., p.109-117 (1997). With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m -2 , far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Z eff . Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δP rad /δZ eff ∝n e 2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful. (orig.)
A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis
Energy Technology Data Exchange (ETDEWEB)
Kadioglu, Samet Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, Ray [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2016-08-01
A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)’s thermal-fluids code) built on top of an other INL’s product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.
International Nuclear Information System (INIS)
Silva Neto, A.J. da; Alvim, A.C.M.
1989-01-01
This work describes the thermalhydraulics code CROSS, designed for micro-computer calculation of heat and mass flow distributions in LWR nuclear reactor cores using the Hardy Cross method. Equations to calculate the pressure variations in the coolant channels are presented, along with derivation of a linear system of equations to calculate the energy balance. This system is solved through the Benachievicz method. A case study is presented, showing that the methodology developed in this work can be used in place of the forward marching multi-channel codes. (author) [pt
International Nuclear Information System (INIS)
Carreira, M.
1965-01-01
As a working method for determination of changes in molecular mass that may occur by irradiation (pyrolytic-radiolytic decomposition) of polyphenyl reactor coolants, a cryoscopic technique has been developed which associated the basic simplicity of Beckman's method with some experimental refinements taken out of the equilibrium methods. A total of 18 runs were made on samples of napthalene, biphenyl, and the commercial mixtures OM-2 (Progil) and Santowax-R (Monsanto), with an average deviation from the theoretical molecular mass of 0.6%. (Author) 7 refs
A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis
International Nuclear Information System (INIS)
Kadioglu, Samet Y.; Berry, Ray; Martineau, Richard
2016-01-01
A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)'s thermal-fluids code) built on top of an other INL's product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.
Homogeneous instantons in bigravity
International Nuclear Information System (INIS)
Zhang, Ying-li; Sasaki, Misao; Yeom, Dong-han
2015-01-01
We study homogeneous gravitational instantons, conventionally called the Hawking-Moss (HM) instantons, in bigravity theory. The HM instantons describe the amplitude of quantum tunneling from a false vacuum to the true vacuum. Corrections to General Relativity (GR) are found in a closed form. Using the result, we discuss the following two issues: reduction to the de Rham-Gabadadze-Tolley (dRGT) massive gravity and the possibility of preference for a large e-folding number in the context of the Hartle-Hawking (HH) no-boundary proposal. In particular, concerning the dRGT limit, it is found that the tunneling through the so-called self-accelerating branch is exponentially suppressed relative to the normal branch, and the probability becomes zero in the dRGT limit. As far as HM instantons are concerned, this could imply that the reduction from bigravity to the dRGT massive gravity is ill-defined.
Analytical solution of point kinetic equations for sub-critical systems
International Nuclear Information System (INIS)
Henrice Junior, Edson; Goncalves, Alessandro C.
2013-01-01
This article presents an analytical solution for the set of point kinetic equations for sub-critical reactors. This solution stems from the ordinary, non-homogeneous differential equation that rules the neutron density and that presents the incomplete Gamma function in its functional form. The method used proved advantageous and allowed practical applications such as the linear insertion of reactivity, considering an external constant source or with both varying linearly. (author)
International Nuclear Information System (INIS)
Ostrovskiy, V.; Kudryavtsev, E.; Tutnov, I.
2011-01-01
The paper presents the basics of approach of planning and carrying out of experiments to validate safety PWR reactors of the future when accepting technical solutions concerning using of improved fuel rods in fuel assembly. Basic principles and criteria used for the validation of technical solutions and developments in improving of nuclear fuel cycle of PWR reactors of the future are presented from the point of safety of future operation of modified fuel rods. We explore the questions of safety operation of PWR reactors with fuel assemblies, containing fuel rods with different length of fuel. The paper discusses the ways of solving of important tasks of critical facility experiments conducting for verification of new technical solutions in the sphere of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations of IEC (International Electromechanical Commission). New Federal laws of Russian Federation define the main principle for demands to NPP and any supplier of nuclear techniques. The principle is 'quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety for any moment of operation of NPP'. On the other hand the second principle should be applied to extraction of the greatest benefit from operation of the equipment, systems or the NPP as whole: 'The long operation and full commercial use of resource and service properties of the equipment, systems and the NPP as a whole'. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components. Solving the practical problems of a validation of safety use of fuel rods with the increased length of a fuel column in fuel assembly in nuclear reactors of the future, we should choose new strategies and programs of verification experiments on the base of the analysis of guarantees
Homogeneous axisymmetric model with a limitting stiff equation of state
International Nuclear Information System (INIS)
Korkina, M.P.; Martynenko, V.G.
1976-01-01
A solution is obtained for Einstein's equations in which all metric coefficients are time functions for a limiting stiff equation of the substance state. Thr solution describes a homogeneous cosmological model with cylindrical symmetry. It is shown that the same metrics can be induced by a massless scalar only time-dependent field. Analysis of this solution is presented
Candu reactors with thorium fuel cycles
International Nuclear Information System (INIS)
Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.
2006-01-01
Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a
International Nuclear Information System (INIS)
Lallement, Dominique.
1979-01-01
Nuclear reactors are monitored by several systems combined. The hydraulic and mechanical limitations on the equipment and the heat transfer requirements in the core set a reliable working range for the boiler defined with certain safety margins. The control system tends to keep the power plant within this working range. The protection system covers all the electrical and mechanical equipment needed to safeguard the boiler in the event of abnormal transients or accidents accounted for in the design of the plant. On units in service protection is handled by cabled automatic systems. For better reliability and safety operation, greater flexibility of use (modularity, adaptability) and improved start-up criteria by data processing the tendency is to use digital programmed systems. Computers are already present in control systems but their introduction into protection systems meets with some reticence on the part of the nuclear safety authorities. A study on the replacement of conventional by digital protection systems is presented. From choices partly made on the principles which should govern the hardware and software of a protection system the reliability of different structures and elements was examined and an experimental model built with its simulator and test facilities. A prototype based on these options and studies is being built and is to be set up on one of the CEN-G reactors for tests [fr
Toward whole-core neutron transport without spatial homogenization
International Nuclear Information System (INIS)
Lewis, E. E.
2009-01-01
Full text of publication follows: A long-term goal of computational reactor physics is the deterministic analysis of power reactor core neutronics without incurring significant discretization errors in the energy, spatial or angular variables. In principle, given large enough parallel configurations with unlimited CPU time and memory, this goal could be achieved using existing three-dimensional neutron transport codes. In practice, however, solving the Boltzmann equation for neutrons over the six-dimensional phase space is made intractable by the nature of neutron cross-sections and the complexity and size of power reactor cores. Tens of thousands of energy groups would be required for faithful cross section representation. Likewise, the numerous material interfaces present in power reactor lattices require exceedingly fine spatial mesh structures; these ubiquitous interfaces preclude effective implementation of adaptive grid, mesh-less methods and related techniques that have been applied so successfully in other areas of engineering science. These challenges notwithstanding, substantial progress continues in the pursuit for more robust deterministic methods for whole-core neutronics analysis. This paper examines the progress over roughly the last decade, emphasizing the space-angle variables and the quest to eliminate errors attributable to spatial homogenization. As prolog we briefly assess 1990's methods used in light water reactor analysis and review the lessons learned from the C5G7 benchmark exercises which were originated in 1999 to appraise the ability of transport codes to perform core calculations without homogenization. We proceed by examining progress over the last decade much of which falls into three areas. These may be broadly characterized as reduced homogenization, dynamic homogenization and planar-axial synthesis. In the first, homogenization in three-dimensional calculations is reduced from the fuel assembly to the pin-cell level. In the second
Energy Technology Data Exchange (ETDEWEB)
Jung, Ki Moon [Korea Institute of Industrial Technology, Cheonan (Korea, Republic of); Choi, Seok Hyun [Key Valve Technologies Ltd., Siheung (Korea, Republic of); Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of)
2017-06-15
Dehydrogenation from the hydrolysis of a sodium borohydride (NaBH{sub 4}) solution has been of interest owing to its high theoretical hydrogen storage capacity (10.8 wt.%) and potentially safe operation. An experimental study has been performed on the catalytic reaction rate and pressure drop of a NaBH4 solution over both a single microchannel with a hydraulic diameter of 300 μm and a staggered array of micro pin fins in the microchannel with hydraulic diameter of 50 μm. The catalytic reaction rates and pressure drops were obtained under Reynolds numbers from 1 to 60 and solution concentrations from 5 to 20 wt.%. Moreover, reacting flows were visualized using a high-speed camera with a macro zoom lens. As a result, both the amount of hydrogenation and pressure drop are 2.45 times and 1.5 times larger in a pin fin microchannel array than in a single microchannel, respectively.
Miller, H.I.; Smith, R.C.
1958-01-21
This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.
Tripathi, Nagesh Kumar; Sathe, Manisha
2017-12-01
Large scale production of activated carbon is need of ongoing research due to its excellent adsorption capacity for removal of heavy metals from contaminated solutions. In the present study, polymeric precursor polystyrene beads [Brunauer Emmett Teller (BET) surface area, 46 m2/g; carbon content, 40.64%; crushing strength, 0.32 kg/sphere] were used to produce a new variant of activated carbon, Activated Carbon Spheres (ACS) in a pilot scale fluidized bed reactor. ACS were prepared by carbonization of polymeric precursor at 850 °C followed by activation of resultant material with steam. Prepared ACS were characterized using scanning electron microscope, CHNS analyzer, thermogravimetric analyzer, surface area analyzer and crushing strength tester. The produced ACS have 1009 m2/g BET surface area, 0.89 cm3/g total pore volume, 92.32% carbon content and 1.1 kg/sphere crushing strength with less than 1% of moisture and ash content. The ACS were also evaluated for its potential to remove hexavalent chromium [Cr(VI)] from contaminated solutions. The chromium removal is observed to be 99.1% at initial concentration 50 mg/l, pH 2, ACS dose 1 g/l, contact time 2 h, agitation 120 rpm and temperature 30 °C. Thus ACS can be used as an adsorbent material for the removal of Cr(VI) from contaminated solutions.
The relationship between continuum homogeneity and statistical homogeneity in cosmology
International Nuclear Information System (INIS)
Stoeger, W.R.; Ellis, G.F.R.; Hellaby, C.
1987-01-01
Although the standard Friedmann-Lemaitre-Robertson-Walker (FLRW) Universe models are based on the concept that the Universe is spatially homogeneous, up to the present time no definition of this concept has been proposed that could in principle be tested by observation. Such a definition is here proposed, based on a simple spatial averaging procedure, which relates observable properties of the Universe to the continuum homogeneity idea that underlies the FLRW models. It turns out that the statistical homogeneity often used to describe the distribution of matter on a large scale does not imply spatial homogeneity according to this definition, and so cannot be simply related to a FLRW Universe model. Values are proposed for the homogeneity parameter and length scale of homogeneity of the Universe. (author)
Homogenization of variational inequalities for obstacle problems
International Nuclear Information System (INIS)
Sandrakov, G V
2005-01-01
Results on the convergence of solutions of variational inequalities for obstacle problems are proved. The variational inequalities are defined by a non-linear monotone operator of the second order with periodic rapidly oscillating coefficients and a sequence of functions characterizing the obstacles. Two-scale and macroscale (homogenized) limiting variational inequalities are obtained. Derivation methods for such inequalities are presented. Connections between the limiting variational inequalities and two-scale and macroscale minimization problems are established in the case of potential operators.
Energy Technology Data Exchange (ETDEWEB)
Teixeira, Paulo Cleber Mendonca
2002-12-01
In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)
International Nuclear Information System (INIS)
Costa, Danilo Leite
2013-01-01
This work aims to present a study about the power distribution behavior in a PWR type reactor, considering both intensity and migration of power peaks due to insertion of control rods into the core. Employing the multidimensional steady-state neutron diffusion equation in order to simulate the neutron flux, and using the Finite Difference Method. Furthermore, based on the axial power distribution on the largest heat flux rod, is carried out thermal analysis of this rod and associated coolant channel. For this purpose is employed the FueLRod 3 D code, it uses the Finite Element Method to model the fuel rod and the associated coolant channel, allowing the thermohydraulics simulation of a single rod discretized in three dimensions, considering the heat flux from the pellet, crossing the gap and the cladding until it reaches the coolant. (author)
Energy Technology Data Exchange (ETDEWEB)
Vallet, Ana, E-mail: avallet@quim.ucm.es [Grupo de Catalisis y Procesos de Separacion (CyPS), Departamento de Ingenieria Quimica, Facultad de Ciencias Quimicas, Universidad Complutense de Madrid, Avda. Complutense s/n, 28040 Madrid (Spain); Besson, Michele, E-mail: michele.besson@ircelyon.univ-lyon1.fr [IRCELYON, Institut de recherches sur la catalyse et l' environnement de Lyon, UMR5256 CNRS-Universite Lyon1, 2 Avenue Albert Einstein, F-69626 Villeurbanne Cedex (France); Ovejero, Gabriel; Garcia, Juan [Grupo de Catalisis y Procesos de Separacion (CyPS), Departamento de Ingenieria Quimica, Facultad de Ciencias Quimicas, Universidad Complutense de Madrid, Avda. Complutense s/n, 28040 Madrid (Spain)
2012-08-15
Highlights: Black-Right-Pointing-Pointer Ni supported over hydrotalcite calcined precursors as catalyst. Black-Right-Pointing-Pointer Catalytic wet air oxidation in trickle bed reactor for Basic Yellow 11 removal. Black-Right-Pointing-Pointer Dye removal depends on temperature, initial dye concentration and flow rate. Black-Right-Pointing-Pointer The catalyst proved to be stable and efficient for the dye degradation. - Abstract: Catalytic wet air oxidation (CWAO) of a Basic Yellow 11 (BY11) aqueous solution, chosen as a model of a hardly biodegradable non-azo dye was carried out in a continuous-flow trickle-bed reactor, using nickel supported over hydrotalcite precursor calcined at 550 Degree-Sign C. An increase in the reaction temperature (120-180 Degree-Sign C), and a decrease in dye concentration (1000-3000 ppm) or liquid flow rate (0.1-0.7 mL min{sup -1}) enhanced the CWAO performance in a 30 and 19% for the variation of the temperature and concentration respectively. After a small leaching observed within the first hours, the catalyst proved to be very stable during the 65-day reaction. The CWAO process was found to be very efficient, achieving BY11 conversion up to 95% and TOC conversion up to 85% at 0.1 mL min{sup -1} and 180 Degree-Sign C under 5 MPa air.
International Nuclear Information System (INIS)
Vallet, Ana; Besson, Michèle; Ovejero, Gabriel; García, Juan
2012-01-01
Highlights: ► Ni supported over hydrotalcite calcined precursors as catalyst. ► Catalytic wet air oxidation in trickle bed reactor for Basic Yellow 11 removal. ► Dye removal depends on temperature, initial dye concentration and flow rate. ► The catalyst proved to be stable and efficient for the dye degradation. - Abstract: Catalytic wet air oxidation (CWAO) of a Basic Yellow 11 (BY11) aqueous solution, chosen as a model of a hardly biodegradable non-azo dye was carried out in a continuous-flow trickle-bed reactor, using nickel supported over hydrotalcite precursor calcined at 550 °C. An increase in the reaction temperature (120–180 °C), and a decrease in dye concentration (1000–3000 ppm) or liquid flow rate (0.1–0.7 mL min −1 ) enhanced the CWAO performance in a 30 and 19% for the variation of the temperature and concentration respectively. After a small leaching observed within the first hours, the catalyst proved to be very stable during the 65-day reaction. The CWAO process was found to be very efficient, achieving BY11 conversion up to 95% and TOC conversion up to 85% at 0.1 mL min −1 and 180 °C under 5 MPa air.
Some properties of spatially homogeneous spacetimes
International Nuclear Information System (INIS)
Coomer, G.C.
1979-01-01
This paper discusses two features of the universe which are influenced in a fundamental way by the spacetime geometry of the universe. The first is the growth of density fluctuations in the early stages of the evolution of the universe. The second is the propagation of electromagnetic radiation in the universe. A spatially homogeneous universe is assumed in both discussions. The gravitational instability theory of galaxy formation is investigated for a viscous fluid and for a charged, conducting fluid with a magnetic field added as a perturbation. It is found that the growth rate of density perturbations in both cases is lower than in the perfect fluid case. Spatially homogeneous but nonisotropic spacetimes are investigated next. Two perfect fluid solutions of Einstein's field equations are found which have spacelike hypersurfaces with Bianchi type II geometry. An expression for the spectrum of the cosmic microwave background radiation in a spatially homogeneous but nonisotropic universe is found. The expression is then used to determine the angular distribution of the intensity of the radiation in the simpler of the two solutions. When accepted values of the matter density and decoupling temperature are inserted into this solution, values for the age of the universe and the time of decoupling are obtained which agree reasonably well with the values of the standard model of the universe
A new formulation for the problem of fuel cell homogenization
International Nuclear Information System (INIS)
Chao, Y.-A.; Martinez, A.S.
1982-01-01
A new homogenization method for reactor cells is described. This new method consists in eliminating the NR approximation for the fuel resonance and the Wigner approximation for the resonance escape probability; the background cross section is then redefined and the problem studied is reanalyzed. (E.G.) [pt
International Nuclear Information System (INIS)
McCoy, D.R.
1981-01-01
S/sub N/ computational benchmark solutions are generated for a onegroup and multigroup fuel-void slab lattice cell which is a rough model of a gas-cooled fast reactor (GCFR) lattice cell. The reactivity induced by the extrusion of the fuel material into the voided region is determined for a series of partially extruded lattice cell configurations. A special modified Gauss S/sub N/ ordinate array design is developed in order to obtain eigenvalues with errors less than 0.03% in all of the configurations that are considered. The modified Gauss S/sub N/ ordinate array design has a substantially improved eigenvalue angular convergence behavior when compared to existing S/sub N/ ordinate array designs used in neutron streaming applications. The angular refinement computations are performed in some cases by using a perturbation theory method which enables one to obtain high order S/sub N/ eigenvalue estimates for greatly reduced computational costs
Diffusion piecewise homogenization via flux discontinuity factors
International Nuclear Information System (INIS)
Sanchez, Richard; Zmijarevic, Igor
2011-01-01
We analyze the calculation of flux discontinuity factors (FDFs) for use with piecewise subdomain assembly homogenization. These coefficients depend on the numerical mesh used to compute the diffusion problem. When the mesh has a single degree of freedom on subdomain interfaces the solution is unique and can be computed independently per subdomain. For all other cases we have implemented an iterative calculation for the FDFs. Our numerical results show that there is no solution to this nonlinear problem but that the iterative algorithm converges towards FDFs values that reproduce subdomains reaction rates with a relatively high precision. In our test we have included both the GET and black-box FDFs. (author)
Directory of Open Access Journals (Sweden)
Ignazio Renato Bellobono
2008-01-01
Full Text Available Photomineralization of methane in air (10.0–1000 ppm (mass/volume of C at 100% relative humidity (dioxygen as oxygen donor was systematically studied at 318±3 K in an annular laboratory-scale reactor by photocatalytic membranes immobilizing titanium dioxide as a function of substrate concentration, absorbed power per unit length of membrane, reactor geometry, and concentration of a proprietary vanadium alkoxide as photopromoter. Kinetics of both substrate disappearance, to yield intermediates, and total organic carbon (TOC disappearance, to yield carbon dioxide, were followed. At a fixed value of irradiance (0.30 W⋅cm-1, the mineralization experiments in gaseous phase were repeated as a function of flow rate (4–400 m3⋅h−1. Moreover, at a standard flow rate of 300 m3⋅h−1, the ratio between the overall reaction volume and the length of the membrane was varied, substantially by varying the volume of reservoir, from and to which circulation of gaseous stream took place. Photomineralization of methane in aqueous solutions was also studied, in the same annular reactor and in the same conditions, but in a concentration range of 0.8–2.0 ppm of C, and by using stoichiometric hydrogen peroxide as an oxygen donor. A kinetic model was employed, from which, by a set of differential equations, four final optimised parameters, k1 and K1, k2 and K2, were calculated, which is able to fit the whole kinetic profile adequately. The influence of irradiance on k1 and k2, as well as of flow rate on K1 and K2, is rationalized. The influence of reactor geometry on k values is discussed in view of standardization procedures of photocatalytic experiments. Modeling of quantum yields, as a function of substrate concentration and irradiance, as well as of concentration of photopromoter, was carried out very satisfactorily. Kinetics of hydroxyl radicals reacting between themselves, leading to hydrogen peroxide, other than with substrate or
Closed-form solution of a two-dimensional fuel temperature model for TRIGA-type reactors
Energy Technology Data Exchange (ETDEWEB)
Rivard, J B [Sandia Laboratories (United States)
1974-07-01
If azimuthal power density variations are ignored, the steady-state temperature distribution within a TRIGA-type fuel element is given by the solution of the Poisson equation in two dimensions (r and z) . This paper presents a closed-form solution of this equation as a function of the axial and radial power density profiles, the conductivity of the U-ZrH, the inlet temperature, specific heat and flow rate of the coolant, and the overall heat transfer coefficient. The method begins with the development of a system of linear ordinary differential equations describing mass and energy balances in the fuel and coolant. From the solution of this system, an expression for the second derivative of the fuel temperature distribution in the axial (z) direction is found. Substitution of this expression into the Poisson equation for T(r,z) reduces it from a partial differential equation to an ordinary differential equation in r, which is subsequently solved in closed-form. The results of typical calculations using the model are presented. (author)
Homogenization of resonant chiral metamaterials
Andryieuski, Andrei; Menzel, Christoph; Rockstuhl, Carsten; Malureanu, Radu; Lederer, Falk; Lavrinenko, Andrei
2010-01-01
Homogenization of metamaterials is a crucial issue as it allows to describe their optical response in terms of effective wave parameters as e.g. propagation constants. In this paper we consider the possible homogenization of chiral metamaterials. We show that for meta-atoms of a certain size a critical density exists above which increasing coupling between neighboring meta-atoms prevails a reasonable homogenization. On the contrary, a dilution in excess will induce features reminiscent to pho...
Bilipschitz embedding of homogeneous fractals
Lü, Fan; Lou, Man-Li; Wen, Zhi-Ying; Xi, Li-Feng
2014-01-01
In this paper, we introduce a class of fractals named homogeneous sets based on some measure versions of homogeneity, uniform perfectness and doubling. This fractal class includes all Ahlfors-David regular sets, but most of them are irregular in the sense that they may have different Hausdorff dimensions and packing dimensions. Using Moran sets as main tool, we study the dimensions, bilipschitz embedding and quasi-Lipschitz equivalence of homogeneous fractals.
International Nuclear Information System (INIS)
Rydell, R.J.
1980-01-01
One objective of this study is to develop a framework of analysis that is useful for investigating the conditions shaping the respective roles of science and politics in decision making on technology policy. The analytical framework used focuses upon the interactive R and D process and specifies the factors affecting change in and of that process. The distinguishing feature of this new analytical framework is its utility for investigating how participants in and R and D process go about defining and solving a growing variety of problems that they encounter as the costs, impacts, and stakes of technological change become more readily apparent. The framework is then applied to a particularly complex and politically controversial technology, the nuclear breeder reactor. Britain and the United States, the original pioneers of technology utilizing plutonium to produce electricity, were singled out in order to test the utility of the analytical framework for the comparative study of the R and D decision-making process. Although the study does not purport to have exhausted all possible interpretations of this complex subject, the results of the study suggest that the interactive R and D process represents an improvement over conventional modes of conceptualizing how R and D policies are formulated and changed. Efforts to resolve major national and international problems relating to science and technology will ultimately succeed only to the extent that these efforts are grounded in a deeper understanding of the conditions affecting how these problems are defined and approached in actual decision-making environments
Energy Technology Data Exchange (ETDEWEB)
Croix, O; Paoli, O; Lecomte, J; Dolle, L; Gallic, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1964-07-01
In the framework of research into the poisoning of the EL-4 reactor by cadmium sulphate, measurements have been made by two different methods of the residual amounts of cadmium liable to be fixed irreversibly on the surfaces in contact with the heavy water. A marked influence of the pH has been noticed. The mechanism of the irreversible fixing is compatible with the hypothesis of an ion-exchange in the surface oxide layer. In a sufficiently wide range of pH the cadmium thus fixed causes very little residual poisoning. The stability of the cadmium sulphate solutions is however rather low in the conditions of poisoning. (authors) [French] Dans le cadre des etudes sur l'empoisonnement du reacteur EL-4 par le sulfate de cadmium, les quantites residuelles de cadmium susceptibles de se fixer irreversiblement sur les parois que mouillerait l'eau lourde, ont ete mesurees experimentalement par deux methodes differentes. On observe une influence nette du pH. Le mecanisme de la fixation irreversible est compatible avec l'hypothese d'un echange d'ions dans la pellicule d'oxyde superficielle. Dans des limites suffisamment larges de pH, la cadmium ainsi fixe n'occasionne pas d'empoisonnement residuel important. La stabilite des solutions de sulfate de cadmium dans les conditions de l'empoisonnement est cependant mediocre. (auteurs)
Estimating minimum polycrystalline aggregate size for macroscopic material homogeneity
International Nuclear Information System (INIS)
Kovac, M.; Simonovski, I.; Cizelj, L.
2002-01-01
During severe accidents the pressure boundary of reactor coolant system can be subjected to extreme loadings, which might cause failure. Reliable estimation of the extreme deformations can be crucial to determine the consequences of severe accidents. Important drawback of classical continuum mechanics is idealization of inhomogenous microstructure of materials. Classical continuum mechanics therefore cannot predict accurately the differences between measured responses of specimens, which are different in size but geometrical similar (size effect). A numerical approach, which models elastic-plastic behavior on mesoscopic level, is proposed to estimate minimum size of polycrystalline aggregate above which it can be considered macroscopically homogeneous. The main idea is to divide continuum into a set of sub-continua. Analysis of macroscopic element is divided into modeling the random grain structure (using Voronoi tessellation and random orientation of crystal lattice) and calculation of strain/stress field. Finite element method is used to obtain numerical solutions of strain and stress fields. The analysis is limited to 2D models.(author)
International Nuclear Information System (INIS)
Cardinali, A.; Morini, L.; Castaldo, C.; Cesario, R.; Zonca, F.
2007-01-01
Knowing that the lower hybrid (LH) wave propagation in tokamak plasmas can be correctly described with a full wave approach only, based on fully numerical techniques or on semianalytical approaches, in this paper, the LH wave equation is asymptotically solved via the Wentzel-Kramers-Brillouin (WKB) method for the first two orders of the expansion parameter, obtaining governing equations for the phase at the lowest and for the amplitude at the next order. The nonlinear partial differential equation (PDE) for the phase is solved in a pseudotoroidal geometry (circular and concentric magnetic surfaces) by the method of characteristics. The associated system of ordinary differential equations for the position and the wavenumber is obtained and analytically solved by choosing an appropriate expansion parameter. The quasilinear PDE for the WKB amplitude is also solved analytically, allowing us to reconstruct the wave electric field inside the plasma. The solution is also obtained numerically and compared with the analytical solution. A discussion of the validity limits of the WKB method is also given on the basis of the obtained results
Reactor core of FBR type reactor
International Nuclear Information System (INIS)
Hayashi, Hideyuki; Ichimiya, Masakazu.
1994-01-01
A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)
Todd, Donald; Tsvetkov, Pavel
2012-01-01
Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...
Homogeneous versus heterogeneous zeolite nucleation
Dokter, W.H.; Garderen, van H.F.; Beelen, T.P.M.; Santen, van R.A.; Bras, W.
1995-01-01
Aggregates of fractal dimension were found in the intermediate gel phases that organize prior to nucleation and crystallization (shown right) of silicalite from a homogeneous reaction mixture. Small- and wide-angle X-ray scattering studies prove that for zeolites nucleation may be homogeneous or
Homogeneous crystal nucleation in polymers.
Schick, C; Androsch, R; Schmelzer, J W P
2017-11-15
The pathway of crystal nucleation significantly influences the structure and properties of semi-crystalline polymers. Crystal nucleation is normally heterogeneous at low supercooling, and homogeneous at high supercooling, of the polymer melt. Homogeneous nucleation in bulk polymers has been, so far, hardly accessible experimentally, and was even doubted to occur at all. This topical review summarizes experimental findings on homogeneous crystal nucleation in polymers. Recently developed fast scanning calorimetry, with cooling and heating rates up to 10 6 K s -1 , allows for detailed investigations of nucleation near and even below the glass transition temperature, including analysis of nuclei stability. As for other materials, the maximum homogeneous nucleation rate for polymers is located close to the glass transition temperature. In the experiments discussed here, it is shown that polymer nucleation is homogeneous at such temperatures. Homogeneous nucleation in polymers is discussed in the framework of the classical nucleation theory. The majority of our observations are consistent with the theory. The discrepancies may guide further research, particularly experiments to progress theoretical development. Progress in the understanding of homogeneous nucleation is much needed, since most of the modelling approaches dealing with polymer crystallization exclusively consider homogeneous nucleation. This is also the basis for advancing theoretical approaches to the much more complex phenomena governing heterogeneous nucleation.
Homogenization of discrete media
International Nuclear Information System (INIS)
Pradel, F.; Sab, K.
1998-01-01
Material such as granular media, beam assembly are easily seen as discrete media. They look like geometrical points linked together thanks to energetic expressions. Our purpose is to extend discrete kinematics to the one of an equivalent continuous material. First we explain how we build the localisation tool for periodic materials according to estimated continuum medium type (classical Cauchy, and Cosserat media). Once the bridge built between discrete and continuum media, we exhibit its application over two bidimensional beam assembly structures : the honey comb and a structural reinforced variation. The new behavior is then applied for the simple plan shear problem in a Cosserat continuum and compared with the real discrete solution. By the mean of this example, we establish the agreement of our new model with real structures. The exposed method has a longer range than mechanics and can be applied to every discrete problems like electromagnetism in which relationship between geometrical points can be summed up by an energetic function. (orig.)
Commensurability effects in holographic homogeneous lattices
International Nuclear Information System (INIS)
Andrade, Tomas; Krikun, Alexander
2016-01-01
An interesting application of the gauge/gravity duality to condensed matter physics is the description of a lattice via breaking translational invariance on the gravity side. By making use of global symmetries, it is possible to do so without scarifying homogeneity of the pertinent bulk solutions, which we thus term as “homogeneous holographic lattices.' Due to their technical simplicity, these configurations have received a great deal of attention in the last few years and have been shown to correctly describe momentum relaxation and hence (finite) DC conductivities. However, it is not clear whether they are able to capture other lattice effects which are of interest in condensed matter. In this paper we investigate this question focusing our attention on the phenomenon of commensurability, which arises when the lattice scale is tuned to be equal to (an integer multiple of) another momentum scale in the system. We do so by studying the formation of spatially modulated phases in various models of homogeneous holographic lattices. Our results indicate that the onset of the instability is controlled by the near horizon geometry, which for insulating solutions does carry information about the lattice. However, we observe no sharp connection between the characteristic momentum of the broken phase and the lattice pitch, which calls into question the applicability of these models to the physics of commensurability.
Coarse mesh finite element method for boiling water reactor physics analysis
International Nuclear Information System (INIS)
Ellison, P.G.
1983-01-01
A coarse mesh method is formulated for the solution of Boiling Water Reactor physics problems using two group diffusion theory. No fuel assembly cross-section homogenization is required; water gaps, control blades and fuel pins of varying enrichments are treated explicitly. The method combines constrained finite element discretization with infinite lattice super cell trial functions to obtain coarse mesh solutions for which the only approximations are along the boundaries between fuel assemblies. The method is applied to bench mark Boiling Water Reactor problems to obtain both the eigenvalue and detailed flux distributions. The solutions to these problems indicate the method is useful in predicting detailed power distributions and eigenvalues for Boiling Water Reactor physics problems
International Nuclear Information System (INIS)
Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko; Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev
2005-01-01
Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m 3 . The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)
Homogenization of discrete media
Energy Technology Data Exchange (ETDEWEB)
Pradel, F.; Sab, K. [CERAM-ENPC, Marne-la-Vallee (France)
1998-11-01
Material such as granular media, beam assembly are easily seen as discrete media. They look like geometrical points linked together thanks to energetic expressions. Our purpose is to extend discrete kinematics to the one of an equivalent continuous material. First we explain how we build the localisation tool for periodic materials according to estimated continuum medium type (classical Cauchy, and Cosserat media). Once the bridge built between discrete and continuum media, we exhibit its application over two bidimensional beam assembly structures : the honey comb and a structural reinforced variation. The new behavior is then applied for the simple plan shear problem in a Cosserat continuum and compared with the real discrete solution. By the mean of this example, we establish the agreement of our new model with real structures. The exposed method has a longer range than mechanics and can be applied to every discrete problems like electromagnetism in which relationship between geometrical points can be summed up by an energetic function. (orig.) 7 refs.
Sashidhar, R B; Selvi, S Kalaignana; Vinod, V T P; Kosuri, Tanuja; Raju, D; Karuna, R
2015-10-01
An ecofriendly green chemistry method using a natural biopolymer, Gum Kondagogu (GK) for the removal of U (VI) from aqueous, simulated nuclear effluents was studied. The adsorption characteristic of GK towards U (VI) from aqueous solution was studied at varied pH, contact time, adsorbent dose, initial U (VI) concentration and temperature using UV-Visible spectroscopy and ICP-MS. Maximum adsorption was seen at pH 4, 0.1% GK with 60 min contact time at room temperature. The GK- U (VI) composite was characterized by FT-IR, zeta potential, TEM and SEM-EDAX. The Langmuir isotherm was found to be 487 mg of U (VI) g(-1) of GK. The adsorption capacity and (%) of U (VI) was found to be 490 ± 5.4 mg g(-1) and 98.5%. Moreover adsorption of U (VI) by GK was not influenced by other cations present in the simulated effluents. The adsorbed U (VI) was efficiently stripped from composite using 1 M HCl. Copyright © 2015 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Sashidhar, R.B.; Selvi, S. Kalaignana; Vinod, V.T.P.; Kosuri, Tanuja; Raju, D.; Karuna, R.
2015-01-01
An ecofriendly green chemistry method using a natural biopolymer, Gum Kondagogu (GK) for the removal of U (VI) from aqueous, simulated nuclear effluents was studied. The adsorption characteristic of GK towards U (VI) from aqueous solution was studied at varied pH, contact time, adsorbent dose, initial U (VI) concentration and temperature using UV–Visible spectroscopy and ICP-MS. Maximum adsorption was seen at pH 4, 0.1% GK with 60 min contact time at room temperature. The GK- U (VI) composite was characterized by FT-IR, zeta potential, TEM and SEM-EDAX. The Langmuir isotherm was found to be 487 mg of U (VI) g −1 of GK. The adsorption capacity and (%) of U (VI) was found to be 490 ± 5.4 mg g −1 and 98.5%. Moreover adsorption of U (VI) by GK was not influenced by other cations present in the simulated effluents. The adsorbed U (VI) was efficiently stripped from composite using 1 M HCl. - Highlights: • An eco-friendly method for removal of U (VI) from simulated nuclear effluents by Gum Kondagogu. • The Langmuir and Freundlich isotherm indicated favourable adsorption. • The adsorption (%) of U (VI) by GK was found to be 98.5%. • Desorption studies on biosorbed metal ions showed that HCl was a good eluent
Verification of homogenization in fast critical assembly analyses
International Nuclear Information System (INIS)
Chiba, Go
2006-01-01
In the present paper, homogenization procedures for fast critical assembly analyses are investigated. Errors caused by homogenizations are evaluated by the exact perturbation theory. In order to obtain reference solutions, three-dimensional plate-wise transport calculations are performed. It is found that the angular neutron flux along plate boundaries has a significant peak in the fission source energy range. To treat this angular dependence accurately, the double-Gaussian Chebyshev angular quadrature set with S 24 is applied. It is shown that the difference between the heterogeneous leakage theory and the homogeneous theory is negligible, and that transport cross sections homogenized with neutron flux significantly underestimate neutron leakage. The error in criticality caused by a homogenization is estimated at about 0.1%Δk/kk' in a small fast critical assembly. In addition, the neutron leakage is overestimated by both leakage theories when sodium plates in fuel lattices are voided. (author)
Homogeneous Spaces and Equivariant Embeddings
Timashev, DA
2011-01-01
Homogeneous spaces of linear algebraic groups lie at the crossroads of algebraic geometry, theory of algebraic groups, classical projective and enumerative geometry, harmonic analysis, and representation theory. By standard reasons of algebraic geometry, in order to solve various problems on a homogeneous space it is natural and helpful to compactify it keeping track of the group action, i.e. to consider equivariant completions or, more generally, open embeddings of a given homogeneous space. Such equivariant embeddings are the subject of this book. We focus on classification of equivariant em
Energy Technology Data Exchange (ETDEWEB)
Tagliaferro, Geronimo V.; Pereira, Paulo Henrique F.; Rodrigues, Liana Alvares; Silva, Maria Lucia Caetano Pinto da, E-mail: fernandes_eng@yahoo.com.b [Universidade de Sao Paulo (USP), Lorena, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Quimica
2011-07-01
This paper describes the adsorption of heavy metals ions from aqueous solution by hydrous niobium oxide. Three heavy metals were selected for this study: cadmium, lead and silver. Adsorption isotherms were well fitted by Langmuir model. Maximum adsorption capacity (Q{sub 0}) for Pb{sup 2+}, Ag{sup +} and Cd{sup 2+} was found to be 452.5, 188.68 and 8.85 mg g{sup -1}, respectively. (author)
Analysis of forced convective transient boiling by homogeneous model of two-phase flow
International Nuclear Information System (INIS)
Kataoka, Isao
1985-01-01
Transient forced convective boiling is of practical importance in relation to the accident analysis of nuclear reactor etc. For large length-to-diameter ratio, the transient boiling characteristics are predicted by transient two-phase flow calculations. Based on homogeneous model of two-phase flow, the transient forced convective boiling for power and flow transients are analysed. Analytical expressions of various parameters of transient two-phase flow have been obtained for several simple cases of power and flow transients. Based on these results, heat flux, velocity and time at transient CHF condition are predicted analytically for step and exponential power increases, and step, exponential and linear velocity decreases. The effects of various parameters on heat flux, velocity and time at transient CHF condition have been clarified. Numerical approach combined with analytical method is proposed for more complicated cases. Solution method for pressure transient are also described. (author)
Qualitative analysis of homogeneous universes
International Nuclear Information System (INIS)
Novello, M.; Araujo, R.A.
1980-01-01
The qualitative behaviour of cosmological models is investigated in two cases: Homogeneous and isotropic Universes containing viscous fluids in a stokesian non-linear regime; Rotating expanding universes in a state which matter is off thermal equilibrium. (Author) [pt
Spinor structures on homogeneous spaces
International Nuclear Information System (INIS)
Lyakhovskii, V.D.; Mudrov, A.I.
1993-01-01
For multidimensional models of the interaction of elementary particles, the problem of constructing and classifying spinor fields on homogeneous spaces is exceptionally important. An algebraic criterion for the existence of spinor structures on homogeneous spaces used in multidimensional models is developed. A method of explicit construction of spinor structures is proposed, and its effectiveness is demonstrated in examples. The results are of particular importance for harmonic decomposition of spinor fields
Vallet, Ana; Besson, Michèle; Ovejero, Gabriel; García, Juan
2012-08-15
Catalytic wet air oxidation (CWAO) of a Basic Yellow 11 (BY11) aqueous solution, chosen as a model of a hardly biodegradable non-azo dye was carried out in a continuous-flow trickle-bed reactor, using nickel supported over hydrotalcite precursor calcined at 550°C. An increase in the reaction temperature (120-180°C), and a decrease in dye concentration (1000-3000 ppm) or liquid flow rate (0.1-0.7 mL min(-1)) enhanced the CWAO performance in a 30 and 19% for the variation of the temperature and concentration respectively. After a small leaching observed within the first hours, the catalyst proved to be very stable during the 65-day reaction. The CWAO process was found to be very efficient, achieving BY11 conversion up to 95% and TOC conversion up to 85% at 0.1 mL min(-1) and 180°C under 5 MPa air. Copyright © 2012 Elsevier B.V. All rights reserved.
The CAREM reactor and present currents in reactor design
International Nuclear Information System (INIS)
Ordonez, J.P.
1990-01-01
INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es
Testing Homogeneity with the Galaxy Fossil Record
Hoyle, Ben; Jimenez, Raul; Heavens, Alan; Clarkson, Chris; Maartens, Roy
2013-01-01
Observationally confirming spatial homogeneity on sufficiently large cosmological scales is of importance to test one of the underpinning assumptions of cosmology, and is also imperative for correctly interpreting dark energy. A challenging aspect of this is that homogeneity must be probed inside our past lightcone, while observations take place on the lightcone. The history of star formation rates (SFH) in the galaxy fossil record provides a novel way to do this. We calculate the SFH of stacked Luminous Red Galaxy (LRG) spectra obtained from the Sloan Digital Sky Survey. We divide the LRG sample into 12 equal area contiguous sky patches and 10 redshift slices (0.2
Molten salt reactors: reactor cores
International Nuclear Information System (INIS)
1983-01-01
In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr
Energy Technology Data Exchange (ETDEWEB)
Magat, Ph
1997-04-01
Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P{sub N} method (SP{sub N}) for the modeling of the core. The comparison between S{sub N} calculations and SP{sub N} calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P{sub 3}, there is no effect; -) the P{sub 1} development is adequate; and -) the P{sub 0} development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP{sub 1}, SP{sub 3}, SP{sub 5} and SP{sub 7}). These calculations show that the implementation of the SP{sub N} method is feasible but the extra-costs in computation times and memory are important. We recommend: SP{sub 5}P{sub 1} calculations for heterogeneous 2-dimension geometry and SP{sub 3}P{sub 1} calculations for the homogeneous 3-dimension geometry. (A.C.)
Energy Technology Data Exchange (ETDEWEB)
Magat, Ph
1997-04-01
Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P{sub N} method (SP{sub N}) for the modeling of the core. The comparison between S{sub N} calculations and SP{sub N} calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P{sub 3}, there is no effect; -) the P{sub 1} development is adequate; and -) the P{sub 0} development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP{sub 1}, SP{sub 3}, SP{sub 5} and SP{sub 7}). These calculations show that the implementation of the SP{sub N} method is feasible but the extra-costs in computation times and memory are important. We recommend: SP{sub 5}P{sub 1} calculations for heterogeneous 2-dimension geometry and SP{sub 3}P{sub 1} calculations for the homogeneous 3-dimension geometry. (A.C.)
7 CFR 58.920 - Homogenization.
2010-01-01
... 7 Agriculture 3 2010-01-01 2010-01-01 false Homogenization. 58.920 Section 58.920 Agriculture... Procedures § 58.920 Homogenization. Where applicable concentrated products shall be homogenized for the... homogenization and the pressure at which homogenization is accomplished will be that which accomplishes the most...
Applications of a systematic homogenization theory for nodal diffusion methods
International Nuclear Information System (INIS)
Zhang, Hong-bin; Dorning, J.J.
1992-01-01
The authors recently have developed a self-consistent and systematic lattice cell and fuel bundle homogenization theory based on a multiple spatial scales asymptotic expansion of the transport equation in the ratio of the mean free path to the reactor characteristics dimension for use with nodal diffusion methods. The mathematical development leads naturally to self-consistent analytical expressions for homogenized diffusion coefficients and cross sections and flux discontinuity factors to be used in nodal diffusion calculations. The expressions for the homogenized nuclear parameters that follow from the systematic homogenization theory (SHT) are different from those for the traditional flux and volume-weighted (FVW) parameters. The calculations summarized here show that the systematic homogenization theory developed recently for nodal diffusion methods yields accurate values for k eff and assembly powers even when compared with the results of a fine mesh transport calculation. Thus, it provides a practical alternative to equivalence theory and GET (Ref. 3) and to simplified equivalence theory, which requires auxiliary fine-mesh calculations for assemblies embedded in a typical environment to determine the discontinuity factors and the equivalent diffusion coefficient for a homogenized assembly
Homogenization of the critically spectral equation in neutron transport
Energy Technology Data Exchange (ETDEWEB)
Allaire, G. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie]|[Paris-6 Univ., 75 (France). Lab. d' Analyse Numerique; Bal, G. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches
1998-07-01
We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)
Homogenization of the critically spectral equation in neutron transport
International Nuclear Information System (INIS)
Allaire, G.; Paris-6 Univ., 75; Bal, G.
1998-01-01
We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)
Computer modeling of homogenization of boric acid in IRIS pressurizer
International Nuclear Information System (INIS)
Rives Sanz, Ronny; Montesinos Otero, Maria Elena; Gonzalez Mantecon, Javier
2015-01-01
Integral layout of nuclear reactor IRIS makes possible the elimination of the spray system; which is usually used to mitigate in-surge transient and help to boron homogenization. The study of transients with deficiencies in the boron homogenization in this technology is very important, because they can cause disturbances in the reactor power and insert a strong reactivity in the core. The aim of the present research is to model the IRIS pressurizer using the CFX code searching for designs alternatives that guaranteed its intrinsic security, focused on the phenomena before mentioned. A symmetric tri dimensional model equivalent to 1/8 of the total geometry was adopted to reduce mesh size and minimize processing time. The relationships are programmed and incorporated into the code. This paper discusses the model developed and the behavior of the system for representative transients sequences. The results of the analyzed IRIS transients could be applied to the design of the pressurizer internal structures and components. (Author)
Note on integrability of certain homogeneous Hamiltonian systems
Energy Technology Data Exchange (ETDEWEB)
Szumiński, Wojciech [Institute of Physics, University of Zielona Góra, Licealna 9, PL-65-407, Zielona Góra (Poland); Maciejewski, Andrzej J. [Institute of Astronomy, University of Zielona Góra, Licealna 9, PL-65-407, Zielona Góra (Poland); Przybylska, Maria, E-mail: M.Przybylska@if.uz.zgora.pl [Institute of Physics, University of Zielona Góra, Licealna 9, PL-65-407, Zielona Góra (Poland)
2015-12-04
In this paper we investigate a class of natural Hamiltonian systems with two degrees of freedom. The kinetic energy depends on coordinates but the system is homogeneous. Thanks to this property it admits, in a general case, a particular solution. Using this solution we derive necessary conditions for the integrability of such systems investigating differential Galois group of variational equations. - Highlights: • Necessary integrability conditions for some 2D homogeneous Hamilton systems are given. • Conditions are obtained analysing differential Galois group of variational equations. • New integrable and superintegrable systems are identified.
Method of the characteristics for calculation of VVER without homogenization
Energy Technology Data Exchange (ETDEWEB)
Suslov, I.R.; Komlev, O.G.; Novikova, N.N.; Zemskov, E.A.; Tormyshev, I.V.; Melnikov, K.G.; Sidorov, E.B. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)
2005-07-01
The first stage of the development of characteristics code MCCG3D for calculation of the VVER-type reactor without homogenization is presented. The parallel version of the code for MPI was developed and tested on cluster PC with LINUX-OS. Further development of the MCCG3D code for design-level calculations with full-scale space-distributed feedbacks is discussed. For validation of the MCCG3D code we use the critical assembly VENUS-2. The geometrical models with and without homogenization have been used. With both models the MCCG3D results agree well with the experimental power distribution and with results generated by the other codes, but model without homogenization provides better results. The perturbation theory for MCCG3D code is developed and implemented in the module KEFSFGG. The calculations with KEFSFGG are in good agreement with direct calculations. (authors)
TWO FERROMAGNETIC SPHERES IN HOMOGENEOUS MAGNETIC FIELD
Directory of Open Access Journals (Sweden)
Yury A. Krasnitsky
2018-01-01
Full Text Available The problem of two spherical conductors is studied quite in detail with bispherical coordinates usage and has numerous appendices in an electrostatics. The boundary-value problem about two ferromagnetic spheres enclosed on homogeneous and infinite environment in which the lack of spheres exists like homogeneous magnetic field is considered. The solution of Laplace's equation in the bispherical system of coordinates allows us to find the potential and field distribution in all spaces, including area between spheres. The boundary conditions in potential continuity and in ordinary density constituent of spheres surfaces induction flux are used. It is supposed that spheres are identical, and magnetic permeability of their material is expressed in >> 0. The problem about falling of electromagnetic plane wave on the system of two spheres, which possesses electrically small sizes, can be considered as quasistationary. The scalar potentials received as a result of Laplace's equation solution are represented by the series containing Legendre polynomials. The concept of two spheres system effective permeability is introduced. It is equal to the advantage in magnitude of magnetic induction flux vector through a certain system’s section arising due to its magnetic properties. Necessary ratios for the effective permeability referred to the central system’s section are obtained. Particularly, the results can be used during the analysis of ferroxcube core clearance, which influences on the magnetic antenna properties.
Genetic Homogenization of Composite Materials
Directory of Open Access Journals (Sweden)
P. Tobola
2009-04-01
Full Text Available The paper is focused on numerical studies of electromagnetic properties of composite materials used for the construction of small airplanes. Discussions concentrate on the genetic homogenization of composite layers and composite layers with a slot. The homogenization is aimed to reduce CPU-time demands of EMC computational models of electrically large airplanes. First, a methodology of creating a 3-dimensional numerical model of a composite material in CST Microwave Studio is proposed focusing on a sufficient accuracy of the model. Second, a proper implementation of a genetic optimization in Matlab is discussed. Third, an association of the optimization script and a simplified 2-dimensional model of the homogeneous equivalent model in Comsol Multiphysics is proposed considering EMC issues. Results of computations are experimentally verified.
International Nuclear Information System (INIS)
Krushelnitsky, V.N.; Berkovich, V.M.; Shvyrayev, Yu.; Podshebaykin, A.K.; Fil, N.S.
2002-01-01
Development of new generation WWER reactors is being carried out in Russia. These new projects with WWER reactors aim to achieve increased levels of safety and reduced costs. This paper describes these designs and discusses the main factors leading to the safety level increase and the improved economics. (author)
Terashima, Y; Ozaki, H; Giri, R R; Tano, T; Nakatsuji, S; Takanami, R; Taniguchi, S
2006-01-01
Environmental pollution by low concentrations of 2,4-Dichlorophenoxyacetic acid (2,4-D) is a concern these days due to ever increasingly stringent regulations. Photocatalysis with immobilized TiO2 fiber is a promising oxidation method. Laboratory experiments on photocatalytic degradation of 0.045 mmol l(-1) 2,4-D with the world's first high-strength TiO2 fiber catalyst were carried out in a continuous flow reactor in which the degradations were, in general, similar to those with high 2,4-D concentrations investigated elsewhere. Degradation and mineralization of 2,4-D were significantly enhanced with no initial pH adjustments. The rate constants for total organic carbon (TOC) without pH adjustment were about two-fold bigger than the pH adjustment cases. CO2 gas measurement and carbon mass-balance were carried out for the first time, where about 34% organic carbon converted into CO2 gas during four-hour oxidation. 2,4-Dichlorophenol (2,4-DCP), phenol, benzyl alcohol and two unknowns (RT = 2.65 and 3.78 min.) were detected as aromatic intermediates while Phenol was the new aromatic in HPLC analysis. Dechlorination efficiencies were high (> 70%) in all the cases, and more than 90% efficiencies were observed in chloride mass balance. Bigger flow rates and solution temperature fixed at 20 degrees C without pH adjustment greatly enhanced 2,4-D mineralization. These results can be an important basis in applying the treatment method for dioxin-contaminated water and wastewater.
Salty popcorn in a homogeneous low-dimensional toy model of holographic QCD
International Nuclear Information System (INIS)
Elliot-Ripley, Matthew
2017-01-01
Recently, a homogeneous ansatz has been used to study cold dense nuclear matter in the Sakai–Sugimoto model of holographic QCD. To justify this homogeneous approximation we here investigate a homogeneous ansatz within a low-dimensional toy version of Sakai–Sugimoto to study finite baryon density configurations and compare it to full numerical solutions. We find the ansatz corresponds to enforcing a dyon salt arrangement in which the soliton solutions are split into half-soliton layers. Within this ansatz we find analogues of the proposed baryonic popcorn transitions, in which solutions split into multiple layers in the holographic direction. The homogeneous results are found to qualitatively match the full numerical solutions, lending confidence to the homogeneous approximations of the full Sakai–Sugimoto model. In addition, we find exact compact solutions in the high density, flat space limit which demonstrate the existence of further popcorn transitions to three layers and beyond. (paper)
International Nuclear Information System (INIS)
Spears, W.R.
1987-01-01
Assuming that the design solutions presently perceived for NET can be extrapolated for use in a power reactor, and using costing experience with present day fusion experiments and with fission power plants, the major components of the cost of a tokamak fusion power reactor are described. The analysis shows the emphasis worth placing on various areas of plant design to reduce costs
Research reactors in Argentina
International Nuclear Information System (INIS)
Carlos Ruben Calabrese
1999-01-01
Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and
A method to calculate spatial xenon oscillations in PWR reactors
International Nuclear Information System (INIS)
Ronig, H.
1976-01-01
The new digital computer programme SEXI for the calculation of spatial Xe oscillations is described. A series expansion of the flux density and the particle densities following the geometrical eigenfunctions of a homogeneous block reactor is chosen as an approach to the solution of the system of differential equations describing this feedback process between neutron flux density and Xe particle density. To calculate the neutron flux density, the time-dependent form of the diffusion equation is used instead of the more common stationary form. Integration is carried out using formal time differential quotients of the Fourier coefficients. (orig./RW) [de
Energy Technology Data Exchange (ETDEWEB)
Ramos Q, R.; Santiago F, C.; Gonzalez P, G., E-mail: ruben.ramos01@cfe.gob.mx [Comision Federal de Electricidad, Central Nuclear Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)
2013-10-15
The nuclear power plant of Laguna Verde carried out the Modernization and Increase of Extended Power Project in its two Units (2005-2011). This modernization included to the electro-hydraulic control system of the main turbine, replacing an ana logical system by one digital (Digital Electro-hydraulic Control - DEHC) whose functions are of controlling the reactor pressure in the different operation ways as wells as of controlling the velocity and load of the main turbine. Also, it has protections that are related with diverse plant systems, as the Reactor Protection Systems (RPS). During the tests stage was realized a programmed load rejection, which Reactor Scram should cause when being presented the shot of main turbine. However, the logic of the RPS was inhibited due to the quick response of the new control DEHC, propitiating a condition of non prospective plant and, in consequence, the Reactor Scram happened for another protection of the RPS. (author)
Bifurcation in the Lengyel–Epstein system for the coupled reactors with diffusion
Directory of Open Access Journals (Sweden)
Shaban Aly
2016-01-01
Full Text Available The main goal of this paper is to continue the investigations of the important system of Fengqi et al. (2008. The occurrence of Turing and Hopf bifurcations in small homogeneous arrays of two coupled reactors via diffusion-linked mass transfer which described by a system of ordinary differential equations is considered. I study the conditions of the existence as well as stability properties of the equilibrium solutions and derive the precise conditions on the parameters to show that the Hopf bifurcation occurs. Analytically I show that a diffusion driven instability occurs at a certain critical value, when the system undergoes a Turing bifurcation, patterns emerge. The spatially homogeneous equilibrium loses its stability and two new spatially non-constant stable equilibria emerge which are asymptotically stable. Numerically, at a certain critical value of diffusion the periodic solution gets destabilized and two new spatially nonconstant periodic solutions arise by Turing bifurcation.
Homogenization models for 2-D grid structures
Banks, H. T.; Cioranescu, D.; Rebnord, D. A.
1992-01-01
In the past several years, we have pursued efforts related to the development of accurate models for the dynamics of flexible structures made of composite materials. Rather than viewing periodicity and sparseness as obstacles to be overcome, we exploit them to our advantage. We consider a variational problem on a domain that has large, periodically distributed holes. Using homogenization techniques we show that the solution to this problem is in some topology 'close' to the solution of a similar problem that holds on a much simpler domain. We study the behavior of the solution of the variational problem as the holes increase in number, but decrease in size in such a way that the total amount of material remains constant. The result is an equation that is in general more complex, but with a domain that is simply connected rather than perforated. We study the limit of the solution as the amount of material goes to zero. This second limit will, in most cases, retrieve much of the simplicity that was lost in the first limit without sacrificing the simplicity of the domain. Finally, we show that these results can be applied to the case of a vibrating Love-Kirchhoff plate with Kelvin-Voigt damping. We rely heavily on earlier results of (Du), (CS) for the static, undamped Love-Kirchhoff equation. Our efforts here result in a modification of those results to include both time dependence and Kelvin-Voigt damping.
Analysis of dynamic stability and safety of reactor system by reactor simulator
International Nuclear Information System (INIS)
Raisic, N.
1963-11-01
In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor
International Nuclear Information System (INIS)
Bussac, J.; Reuss, P.
1985-01-01
This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr
Resonance integral calculations for high temperature reactors
International Nuclear Information System (INIS)
Blake, J.P.H.
1960-02-01
Methods of calculation of resonance integrals of finite dilution and temperature are given for both, homogeneous and heterogeneous geometries, together with results obtained from these methods as applied to the design of high temperature reactors. (author)
Size-dependent homogenized diffusion parameters for a finite lattice
International Nuclear Information System (INIS)
Premuda, F.
1980-01-01
A numerical technique is reported for solving the transcendental equation for unknown Ysub(n+1). The solution is expressed in terms of quantities related to Ysub(n). This is an iterative reversion technique which has already been proven to converge rapidly in the homogeneous slab problem considered herein. (author)
Non-homogeneous polymer model for wave propagation and its ...
African Journals Online (AJOL)
user
density are functions of space i.e. non-homogeneous engineering material. .... The Solution of equation Eq. (9) in the form of Eq. (10) can be obtained by taking a phase ..... Viscoelastic Model Applied to a Particular Case .... p m i exp m α α σ σ σ. = −. +. −. (35). The progressive harmonic wave which starts from the end. 0 x =.
Observational homogeneity of the Universe
International Nuclear Information System (INIS)
Bonnor, W.B.; Ellis, G.F.R.
1986-01-01
A new approach to observational homogeneity is presented. The observation that stars and galaxies in distant regions appear similar to those nearby may be taken to imply that matter has had a similar thermodynamic history in widely separated parts of the Universe (the Postulate of Uniform Thermal Histories, or PUTH). The supposition is now made that similar thermodynamic histories imply similar dynamical histories. Then the distant apparent similarity is evidence for spatial homogeneity of the Universe. General Relativity is used to test this idea, taking a perfect fluid model and implementing PUTH by the condition that the density and entropy per baryon shall be the same function of the proper time along all galaxy world-lines. (author)
Conclusions about homogeneity and devitrification
International Nuclear Information System (INIS)
Larche, F.
1997-01-01
A lot of experimental data concerning homogeneity and devitrification of R7T7 glass have been published. It appears that: - the crystallization process is very limited, - the interfaces due to bubbles and the container wall favor crystallization locally but the ratio of crystallized volume remains always below a few per cents, and - crystallization has no damaging long-term effects as far as leaching tests can be trusted. (A.C.)
Is charity a homogeneous good?
Backus, Peter
2010-01-01
In this paper I estimate income and price elasticities of donations to six different charitable causes to test the assumption that charity is a homogeneous good. In the US, charitable donations can be deducted from taxable income. This has long been recognized as producing a price, or taxprice, of giving equal to one minus the marginal tax rate faced by the donor. A substantial portion of the economic literature on giving has focused on estimating price and income elasticities of giving as th...
International Nuclear Information System (INIS)
Naudan, G.; Nigon, J.L.
1993-01-01
After principles of chain reaction and criticality notion, a descriptive model of neutrons behaviour is exposed from a local point of view (this model is called four factors model). One justifies the use of middle values for the calculation of the distribution in space of reactor, quantities representing heterogeneous middle from a local point of view (fuel, moderator, can or clad, and so on ...) by substitution of an equivalent homogeneous middle. Time dependence, dynamical behaviour of reactor are studied. Long term effects of evolution of constituents elements of heart under irradiation, and ways to balance this evolution are in the last paragraph. 18 refs., 26 figs
Actinide recycling in reactors
International Nuclear Information System (INIS)
Kuesters, H.; Wiese, H.W.; Krieg, B.
1995-01-01
The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of 'risk nuclides' in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)
Effect of homogeneous and heterogeneous reactions on the solute ...
African Journals Online (AJOL)
user
and mass transfer flow past a vertical porous flat plate, in the presence of heat ... 2000) due to practical applications, including chemical engineering, soil science and ... a porous medium and a fluid layer, including the inertia and boundary.
Affine Toda equations and solutions in the homogeneous grading
Czech Academy of Sciences Publication Activity Database
Zuevsky, Alexander
2018-01-01
Roč. 542, April 1 (2018), s. 149-161 ISSN 0024-3795 Institutional support: RVO:67985840 Keywords : affine Lie gebras * affine Toda modes * solitons Subject RIV: BA - General Mathematics OBOR OECD: Pure mathematics Impact factor: 0.973, year: 2016 https://www.sciencedirect.com/science/article/pii/S0024379517302100
International Nuclear Information System (INIS)
Barre, Bertrand
2015-10-01
After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor
International Nuclear Information System (INIS)
Vorona, P.M.; Razbudej, V.F.
2010-01-01
Calculational studies and analysis of the neutron fields of WWR-M research reactor of the Institute for Nuclear Research, National Academy of Sciences of Ukraine, as a basic nuclear facility for performing the fundamental and applied investigations and for experimentalindustrial production of radioisotope products for various spheres of application are carried out. The calculations are carried out by the method of statistic tests (Monte Carlo) applying the computer program MCNP-4C. The data on the spectra and the neutron flux density values at the 10 MW reactor power for all technological facilities designed for the works with neutrons: 19 vertical experimental channels for irradiation of specimens and 10 horizontal channels for beams extraction from the reactor are obtained. The effect of the neutron traps (water cavities) mounted in the core on the characteristics of the extracted from the reactor beams is demonstrated. Recommendations associated with optimization of the reactor core are adduced for amplification of its capabilities as a neutron source in experimental researches.
Fourier-Accelerated Nodal Solvers (FANS) for homogenization problems
Leuschner, Matthias; Fritzen, Felix
2017-11-01
Fourier-based homogenization schemes are useful to analyze heterogeneous microstructures represented by 2D or 3D image data. These iterative schemes involve discrete periodic convolutions with global ansatz functions (mostly fundamental solutions). The convolutions are efficiently computed using the fast Fourier transform. FANS operates on nodal variables on regular grids and converges to finite element solutions. Compared to established Fourier-based methods, the number of convolutions is reduced by FANS. Additionally, fast iterations are possible by assembling the stiffness matrix. Due to the related memory requirement, the method is best suited for medium-sized problems. A comparative study involving established Fourier-based homogenization schemes is conducted for a thermal benchmark problem with a closed-form solution. Detailed technical and algorithmic descriptions are given for all methods considered in the comparison. Furthermore, many numerical examples focusing on convergence properties for both thermal and mechanical problems, including also plasticity, are presented.
Physical applications of homogeneous balls
Scarr, Tzvi
2005-01-01
One of the mathematical challenges of modern physics lies in the development of new tools to efficiently describe different branches of physics within one mathematical framework. This text introduces precisely such a broad mathematical model, one that gives a clear geometric expression of the symmetry of physical laws and is entirely determined by that symmetry. The first three chapters discuss the occurrence of bounded symmetric domains (BSDs) or homogeneous balls and their algebraic structure in physics. The book further provides a discussion of how to obtain a triple algebraic structure ass
Heterotic strings on homogeneous spaces
International Nuclear Information System (INIS)
Israel, D.; Kounnas, C.; Orlando, D.; Petropoulos, P.M.
2005-01-01
We construct heterotic string backgrounds corresponding to families of homogeneous spaces as exact conformal field theories. They contain left cosets of compact groups by their maximal tori supported by NS-NS 2-forms and gauge field fluxes. We give the general formalism and modular-invariant partition functions, then we consider some examples such as SU(2)/U(1)∝S 2 (already described in a previous paper) and the SU(3)/U(1) 2 flag space. As an application we construct new supersymmetric string vacua with magnetic fluxes and a linear dilaton. (Abstract Copyright [2005], Wiley Periodicals, Inc.)
Homogenization of variational inequalities and equations defined by pseudomonotone operators
International Nuclear Information System (INIS)
Sandrakov, G V
2008-01-01
Results on the convergence of sequences of solutions of non-linear equations and variational inequalities for obstacle problems are proved. The variational inequalities and equations are defined by a non-linear, pseudomonotone operator of the second order with periodic, rapidly oscillating coefficients and by sequences of functions characterizing the obstacles and the boundary conditions. Two-scale and macroscale (homogenized) limiting problems for such variational inequalities and equations are obtained. Results on the relationship between solutions of these limiting problems are established and sufficient conditions for the uniqueness of solutions are presented. Bibliography: 25 titles
Energy Technology Data Exchange (ETDEWEB)
Cadilhac, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1963-11-15
After a general survey of the theory of neutron thermalization in homogeneous media, one introduces, through a proper formulation, a simplified model generalizing both the Horowitz model (generalized heavy free gas approximation) and the proton gas model. When this model is used, the calculation of spectra is reduced to the solution of linear second order differential equations. Since it depends on two arbitrary functions, the model gives a good approximation of any usual moderator for reactor physics purposes. The choice of these functions is discussed from a theoretical point of view; a method based on the consideration of the first two moments of the scattering law is investigated. Finally, the possibility of discriminating models by using experimental informations is considered. (author) [French] Apres un passage en revue de generalites sur la thermalisation des neutrons dans les milieux homogenes, on developpe un formalisme permettant de definir et d'etudier un modele simplifie de thermaliseur. Ce modele generalise l'approximation proposee par J. HOROWITZ (''gaz lourd generalise'') et comporte comme cas particulier le modele ''hydrogene gazeux monoatomique''. Il ramene le calcul des spectres a la resolution d'equations differentielles lineaires du second ordre. Il fait intervenir deux fonctions arbitraires, ce qui lui permet de representer les thermaliseurs usuels de facon satisfaisante pour les besoins de la physique des reacteurs. L'ajustement theorique de ces fonctions est discute; on etudie une methode basee sur la consideration des deux premiers moments de la loi de diffusion. On envisage enfin la possibilite de discriminer les modeles d'apres des renseignements d'origine experimentale. (auteur)
Homogenization scheme for acoustic metamaterials
Yang, Min
2014-02-26
We present a homogenization scheme for acoustic metamaterials that is based on reproducing the lowest orders of scattering amplitudes from a finite volume of metamaterials. This approach is noted to differ significantly from that of coherent potential approximation, which is based on adjusting the effective-medium parameters to minimize scatterings in the long-wavelength limit. With the aid of metamaterials’ eigenstates, the effective parameters, such as mass density and elastic modulus can be obtained by matching the surface responses of a metamaterial\\'s structural unit cell with a piece of homogenized material. From the Green\\'s theorem applied to the exterior domain problem, matching the surface responses is noted to be the same as reproducing the scattering amplitudes. We verify our scheme by applying it to three different examples: a layered lattice, a two-dimensional hexagonal lattice, and a decorated-membrane system. It is shown that the predicted characteristics and wave fields agree almost exactly with numerical simulations and experiments and the scheme\\'s validity is constrained by the number of dominant surface multipoles instead of the usual long-wavelength assumption. In particular, the validity extends to the full band in one dimension and to regimes near the boundaries of the Brillouin zone in two dimensions.
ISOTOPE METHODS IN HOMOGENEOUS CATALYSIS.
Energy Technology Data Exchange (ETDEWEB)
BULLOCK,R.M.; BENDER,B.R.
2000-12-01
The use of isotope labels has had a fundamentally important role in the determination of mechanisms of homogeneously catalyzed reactions. Mechanistic data is valuable since it can assist in the design and rational improvement of homogeneous catalysts. There are several ways to use isotopes in mechanistic chemistry. Isotopes can be introduced into controlled experiments and followed where they go or don't go; in this way, Libby, Calvin, Taube and others used isotopes to elucidate mechanistic pathways for very different, yet important chemistries. Another important isotope method is the study of kinetic isotope effects (KIEs) and equilibrium isotope effect (EIEs). Here the mere observation of where a label winds up is no longer enough - what matters is how much slower (or faster) a labeled molecule reacts than the unlabeled material. The most careti studies essentially involve the measurement of isotope fractionation between a reference ground state and the transition state. Thus kinetic isotope effects provide unique data unavailable from other methods, since information about the transition state of a reaction is obtained. Because getting an experimental glimpse of transition states is really tantamount to understanding catalysis, kinetic isotope effects are very powerful.
Electrical model of dielectric barrier discharge homogenous and filamentary modes
López-Fernandez, J. A.; Peña-Eguiluz, R.; López-Callejas, R.; Mercado-Cabrera, A.; Valencia-Alvarado, R.; Muñoz-Castro, A.; Rodríguez-Méndez, B. G.
2017-01-01
This work proposes an electrical model that combines homogeneous and filamentary modes of an atmospheric pressure dielectric barrier discharge cell. A voltage controlled electric current source has been utilized to implement the power law equation that represents the homogeneous discharge mode, which starts when the gas breakdown voltage is reached. The filamentary mode implies the emergence of electric current conducting channels (microdischarges), to add this phenomenon an RC circuit commutated by an ideal switch has been proposed. The switch activation occurs at a higher voltage level than the gas breakdown voltage because it is necessary to impose a huge electric field that contributes to the appearance of streamers. The model allows the estimation of several electric parameters inside the reactor that cannot be measured. Also, it is possible to appreciate the modes of the DBD depending on the applied voltage magnitude. Finally, it has been recognized a good agreement between simulation outcomes and experimental results.
Neutron transport equation - indications on homogenization and neutron diffusion
International Nuclear Information System (INIS)
Argaud, J.P.
1992-06-01
In PWR nuclear reactor, the practical study of the neutrons in the core uses diffusion equation to describe the problem. On the other hand, the most correct method to describe these neutrons is to use the Boltzmann equation, or neutron transport equation. In this paper, we give some theoretical indications to obtain a diffusion equation from the general transport equation, with some simplifying hypothesis. The work is organised as follows: (a) the most general formulations of the transport equation are presented: integro-differential equation and integral equation; (b) the theoretical approximation of this Boltzmann equation by a diffusion equation is introduced, by the way of asymptotic developments; (c) practical homogenization methods of transport equation is then presented. In particular, the relationships with some general and useful methods in neutronic are shown, and some homogenization methods in energy and space are indicated. A lot of other points of view or complements are detailed in the text or the remarks
Electrical model of dielectric barrier discharge homogenous and filamentary modes
International Nuclear Information System (INIS)
López-Fernandez, J A; Peña-Eguiluz, R; López-Callejas, R; Mercado-Cabrera, A; Valencia-Alvarado, R; Muñoz-Castro, A; Rodríguez-Méndez, B G
2017-01-01
This work proposes an electrical model that combines homogeneous and filamentary modes of an atmospheric pressure dielectric barrier discharge cell. A voltage controlled electric current source has been utilized to implement the power law equation that represents the homogeneous discharge mode, which starts when the gas breakdown voltage is reached. The filamentary mode implies the emergence of electric current conducting channels (microdischarges), to add this phenomenon an RC circuit commutated by an ideal switch has been proposed. The switch activation occurs at a higher voltage level than the gas breakdown voltage because it is necessary to impose a huge electric field that contributes to the appearance of streamers. The model allows the estimation of several electric parameters inside the reactor that cannot be measured. Also, it is possible to appreciate the modes of the DBD depending on the applied voltage magnitude. Finally, it has been recognized a good agreement between simulation outcomes and experimental results. (paper)
Versteeg, G.F.; Swaaij, W.P.M. van
1988-01-01
Absorption rates of H2S and CO2 in several aqueous alkanolamines in a cocurrent downflow fixed-bed reactor operated in the pulse flow regime have been measured in order to obtain information on the potential selectivity and on the mass transfer parameters. From these experiments it can be concluded
Calculation models for a nuclear reactor
International Nuclear Information System (INIS)
Tashanii, Ahmed Ali
2010-01-01
Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)
Homogeneous Minor Actinide Transmutation in SFR: Neutronic Uncertainties Propagation with Depletion
International Nuclear Information System (INIS)
Buiron, L.; Plisson-Rieunier, D.
2015-01-01
In the frame of next generation fast reactor design, the minimisation of nuclear waste production is one of the key objectives for current R and D. Among the possibilities studied at CEA, minor actinides multi-recycling is the most promising industrial way achievable in the near-term. Two main management options are considered: - Multi-recycling in a homogeneous way (minor actinides diluted in the driver fuel). If this solution can help achieving high transmutation rates, the negative impact of minor actinides on safety coefficients allows only a small fraction of the total heavy mass to be loaded in the core (∼ few %). - Multi-recycling in heterogeneous way by means of Minor Actinide Bearing Blanket (MABB) located at the core periphery. This solution offers more flexibility than the previous one, allowing a total minor actinides decoupled management from the core fuel. As the impact on feedback coefficient is small larger initial minor actinide mass can be loaded in this configuration. Starting from a breakeven Sodium Fast Reactor designed jointly by CEA, Areva and EdF teams, the so called SFR V2B, transmutation performances have been studied in frame on the French fleet for both options and various specific isotopic management (all minor actinides, americium only, etc.). Using these results, a sensitivity study has been performed to assess neutronic uncertainties (i.e coming from cross section) on mass balance on the most attractive configurations. This work in based on a new implementation of sensitivity on concentration with depletion in the ERANOS code package. Uncertainties on isotopes masses at the end of irradiation using various variance-covariance is discussed. (authors)
Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...
Improving homogeneity by dynamic speed limit systems.
Nes, N. van Brandenberg, S. & Twisk, D.A.M.
2010-01-01
Homogeneity of driving speeds is an important variable in determining road safety; more homogeneous driving speeds increase road safety. This study investigates the effect of introducing dynamic speed limit systems on homogeneity of driving speeds. A total of 46 subjects twice drove a route along 12
7 CFR 58.636 - Homogenization.
2010-01-01
... 7 Agriculture 3 2010-01-01 2010-01-01 false Homogenization. 58.636 Section 58.636 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Standards... Procedures § 58.636 Homogenization. Homogenization of the pasteurized mix shall be accomplished to...
The homogeneous geometries of real hyperbolic space
DEFF Research Database (Denmark)
Castrillón López, Marco; Gadea, Pedro Martínez; Swann, Andrew Francis
We describe the holonomy algebras of all canonical connections of homogeneous structures on real hyperbolic spaces in all dimensions. The structural results obtained then lead to a determination of the types, in the sense of Tricerri and Vanhecke, of the corresponding homogeneous tensors. We use...... our analysis to show that the moduli space of homogeneous structures on real hyperbolic space has two connected components....
Orthogonality Measurement for Homogenous Projects-Bases
Ivan, Ion; Sandu, Andrei; Popa, Marius
2009-01-01
The homogenous projects-base concept is defined. Next, the necessary steps to create a homogenous projects-base are presented. A metric system is built, which then will be used for analyzing projects. The indicators which are meaningful for analyzing a homogenous projects-base are selected. The given hypothesis is experimentally verified. The…
International Nuclear Information System (INIS)
Shen, W.
2012-01-01
Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)
Energy Technology Data Exchange (ETDEWEB)
Shen, W. [Candu Energy Inc., 2285 Speakman Dr., Mississauga, ON L5B 1K (Canada)
2012-07-01
Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)
The analytical solution to the 1D diffusion equation in heterogeneous media
International Nuclear Information System (INIS)
Ganapol, B.D.; Nigg, D.W.
2011-01-01
The analytical solution to the time-independent multigroup diffusion equation in heterogeneous plane cylindrical and spherical media is presented. The solution features the simplicity of the one-group formulation while addressing the complication of multigroup diffusion in a fully heterogeneous medium. Beginning with the vector form of the diffusion equation, the approach, based on straightforward mathematics, resolves a set of coupled second order ODEs. The analytical form is facilitated through matrix diagonalization of the neutron interaction matrix rendering the multigroup solution as a series of one-group solutions which, when re-assembled, gives the analytical solution. Customized Eigenmode solutions of the one-group diffusion operator then represent the homogeneous solution in a uniform spatial domain. Once the homogeneous solution is known, the particular solution naturally emerges through variation of parameters. The analytical expression is then numerically implemented through recurrence. Finally, we apply the theory to assess the accuracy of a second order finite difference scheme and to a 1D slab BWR reactor in the four-group approximation. (author)
Preliminary development of thermal nuclear cell homogenization code
International Nuclear Information System (INIS)
Su'ud, Z.; Shafii, M. A.; Yudha, S. P.; Waris, A.; Rijal, K.
2012-01-01
Nuclear fuel cell homogenization for thermal reactors usually include three main parts, i.e., fast energy resonance part which usually adopt narrow resonance approximation to treat the resonance, low (intermediate) energy region in which the resonance can not be treated accurately using NR approximation and therefore we should use intermediate resonance treatment, and thermal energy region (very low) in which the effect of thermal must be treated properly. In n this study the application of the intermediate resonance approximation treatment for low energy nuclear resonance is discussed. The method is iterative based. As a sample the method is applied in U-235 low lying resonance and the result is presented and discussed.
Modeling the homogenization kinetics of as-cast U-10wt% Mo alloys
Energy Technology Data Exchange (ETDEWEB)
Xu, Zhijie, E-mail: zhijie.xu@pnnl.gov [Computational Mathematics Group, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Joshi, Vineet [Energy Processes & Materials Division, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Hu, Shenyang [Reactor Materials & Mechanical Design, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Paxton, Dean [Nuclear Engineering and Analysis Group, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Lavender, Curt [Energy Processes & Materials Division, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Burkes, Douglas [Nuclear Engineering and Analysis Group, Pacific Northwest National Laboratory, Richland, WA 99352 (United States)
2016-04-01
Low-enriched U-22at% Mo (U–10Mo) alloy has been considered as an alternative material to replace the highly enriched fuels in research reactors. For the U–10Mo to work effectively and replace the existing fuel material, a thorough understanding of the microstructure development from as-cast to the final formed structure is required. The as-cast microstructure typically resembles an inhomogeneous microstructure with regions containing molybdenum-rich and -lean regions, which may affect the processing and possibly the in-reactor performance. This as-cast structure must be homogenized by thermal treatment to produce a uniform Mo distribution. The development of a modeling capability will improve the understanding of the effect of initial microstructures on the Mo homogenization kinetics. In the current work, we investigated the effect of as-cast microstructure on the homogenization kinetics. The kinetics of the homogenization was modeled based on a rigorous algorithm that relates the line scan data of Mo concentration to the gray scale in energy dispersive spectroscopy images, which was used to generate a reconstructed Mo concentration map. The map was then used as realistic microstructure input for physics-based homogenization models, where the entire homogenization kinetics can be simulated and validated against the available experiment data at different homogenization times and temperatures.
Khosravi, Morteza; Rakhshaee, Roohan; Ganji, Masuod Taghi
2005-12-09
Intact and treated biomass can remove heavy metals from water and wastewater. This study examined the ability of the activated, semi-intact and inactivated Azolla filiculoides (a small water fern) to remove Pb(2+), Cd(2+), Ni(2+) and Zn(2+) from the aqueous solution. The maximum uptake capacities of these metal ions using the activated Azolla filiculoides by NaOH at pH 10.5 +/- 0.2 and then CaCl(2)/MgCl(2)/NaCl with total concentration of 2 M (2:1:1 mole ratio) in the separate batch reactors were obtained about 271, 111, 71 and 60 mg/g (dry Azolla), respectively. The obtained capacities of maximum adsorption for these kinds of the pre-treated Azolla in the fixed-bed reactors (N(o)) were also very close to the values obtained for the batch reactors (Q(max)). On the other hand, it was shown that HCl, CH(3)OH, C(2)H(5)OH, FeCl(2), SrCl(2), BaCl(2) and AlCl(3) in the pre-treatment processes decreased the ability of Azolla to remove the heavy metals in comparison to the semi-intact Azolla, considerably. The kinetic studies showed that the heavy metals uptake by the activated Azolla was done more rapid than those for the semi-intact Azolla.
The evaporative vector: Homogeneous systems
International Nuclear Information System (INIS)
Klots, C.E.
1987-05-01
Molecular beams of van der Waals molecules are the subject of much current research. Among the methods used to form these beams, three-sputtering, laser ablation, and the sonic nozzle expansion of neat gases - yield what are now recognized to be ''warm clusters.'' They contain enough internal energy to undergo a number of first-order processes, in particular that of evaporation. Because of this evaporation and its attendant cooling, the properties of such clusters are time-dependent. The states of matter which can be arrived at via an evaporative vector on a typical laboratory time-scale are discussed. Topics include the (1) temperatures, (2) metastability, (3) phase transitions, (4) kinetic energies of fragmentation, and (5) the expression of magical properties, all for evaporating homogeneous clusters
Homogenization of Large-Scale Movement Models in Ecology
Garlick, M.J.; Powell, J.A.; Hooten, M.B.; McFarlane, L.R.
2011-01-01
A difficulty in using diffusion models to predict large scale animal population dispersal is that individuals move differently based on local information (as opposed to gradients) in differing habitat types. This can be accommodated by using ecological diffusion. However, real environments are often spatially complex, limiting application of a direct approach. Homogenization for partial differential equations has long been applied to Fickian diffusion (in which average individual movement is organized along gradients of habitat and population density). We derive a homogenization procedure for ecological diffusion and apply it to a simple model for chronic wasting disease in mule deer. Homogenization allows us to determine the impact of small scale (10-100 m) habitat variability on large scale (10-100 km) movement. The procedure generates asymptotic equations for solutions on the large scale with parameters defined by small-scale variation. The simplicity of this homogenization procedure is striking when compared to the multi-dimensional homogenization procedure for Fickian diffusion,and the method will be equally straightforward for more complex models. ?? 2010 Society for Mathematical Biology.
Homogenization of compacted blends of Ni and Mo powders
International Nuclear Information System (INIS)
Lanam, R.D.; Yeh, F.C.H.; Rovsek, J.E.; Smith, D.W.; Heckel, R.W.
1975-01-01
The homogenization behavior of compacted blends of Ni and Mo powders was studied primarily as a function of temperature, mean compact composition, and Mo powder particle size. All compact compositions were in the Ni-rich terminal solid-solution range; temperatures were between 950 and 1200 0 C (in the region of the phase diagram where only the Mo--Ni intermediate phase forms); average Mo particle sizes ranged from 8.4 mu m to 48 mu m. Homogenization was characterized in terms of the rate of decrease of the amounts of the Mo-rich terminal solid-solution phase and the Mo--Ni intermediate phase. The experimental results were compared to predictions based upon the three-phase, concentric-sphere homogenization model. In general, agreement between experimental data and model predictions was fairly good for high-temperature treatments and for compact compositions which were not close to the solubility limit of Mo in Ni. Departures from the model are discussed in terms of surface diffusion contributions to homogenization and non-uniform mixing effects. (U.S.)
International Nuclear Information System (INIS)
Ingremeau, J.-J.X.
2011-01-01
In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc...) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) uses analytical models and interpolations (Meta-models) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a 'buffer', a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that
International Nuclear Information System (INIS)
Rahnema, Farzad
2009-01-01
This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based solely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.
Kinetics of two phase fuel reflected reactors
International Nuclear Information System (INIS)
Buzano, M.L.; Corno, S.E.; Mattioda, F.
2000-01-01
In the present work a self-consistent mathematical model for the local dynamics of a quite particular class of fission reactors has been developed and solved. These devices consist of an innermost multiplying region, in which a significant fraction of the fissile fuel is diluted into a liquid phase, while the complementary fuel fraction operates as a standing solid matrix. This unconventional active region is surrounded by a standard peripheral reflector. For cooling purposes, the fluid fraction of the fuel needs to be circulated through external heat exchangers. The pump driven circulation causes the delayed neutron precursors, dissolved inside the fluid phase, to be spatially homogenized in the core volume well before decaying, while a continuous removal of precursor nuclei from the core takes place as a consequence of the outside circulation. Furthermore, the fraction of the extracted precursors still surviving after the solenoidal trip through the heat exchangers is continuously reinserted into the core. A new type of dynamical model is required to account for these unusual technological features. The mathematical structure of the evolution model presented in this paper consists of a system of integro-differential-difference equations, whose solution is derived in closed-form, by means of fully analytical techniques. Many dynamics and safety features of reactors of this type can be clarified a priori, upon inspection of the mathematical properties of the solution of the model. The rigorous time-eigenvalue generating equation can be explicitly established in the present theoretical context, together with the evaluation of any kind of transients. A short survey on the possible fields of application of these reactors is also presented
The development of a transient neutron flux solution in the PANTHER code
International Nuclear Information System (INIS)
Hutt, P.K.; Knight, M.P.
1990-01-01
In the United Kingdom a new three-dimensional, two-group, homogeneous reactor diffusion code, PANTHER, has been developed for the analysis of pressurized water reactors (PWRs) and advanced gas-cooled reactors (AGRs). The code can perform a comprehensive range of calculations, steady state, depletion, and transient with either a finite difference or analytic nodal flux solution. The nodal solution allows the representation of within-node burnup variation and pin-power reconstruction in either steady-state or transient mode. Specific steady-state and transient thermal feedback modules are included for both PWRs and AGRs. The code is being developed to perform a complete range of reactor calculations from online operational support to fuel management and fault transient analysis. In the area of transient analysis, the code is currently being used for a number of PWR fault transient assessments, including rod ejection and steam-line break. In addition, work is proceeding to incorporate the PANTHER 3D nodal transient solution in the TRAC-P code. This paper outlines the development of the transient flux solutions within PANTHER
Reactivity feedback components of a homogeneous U10Zr-fueled 900 MWt LMR
International Nuclear Information System (INIS)
Meneghetti, D.; Kucera, D.A.
1988-01-01
The linear and Doppler feedback components of the regional contributions of the power-reactivity-decrement (PRD) and of the temperature coefficient of reactivity for a 900 MWt homogeneous U10Zr-fueled sodium-cooled reactor are calculated. The PRD components are separated into power dependent and power-to-flow dependent parts. The values of PRD and temperature coefficient components are compared with corresponding quantities calculated for the Experimental Breeder Reactor II. The implications of these comparisons upon inherent safety characteristics of metal-fueled sodium-cooled reactors are discussed
International Nuclear Information System (INIS)
Sood, D.D.
1980-01-01
Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)
Reciprocity theory of homogeneous reactions
Agbormbai, Adolf A.
1990-03-01
The reciprocity formalism is applied to the homogeneous gaseous reactions in which the structure of the participating molecules changes upon collision with one another, resulting in a change in the composition of the gas. The approach is applied to various classes of dissociation, recombination, rearrangement, ionizing, and photochemical reactions. It is shown that for the principle of reciprocity to be satisfied it is necessary that all chemical reactions exist in complementary pairs which consist of the forward and backward reactions. The backward reaction may be described by either the reverse or inverse process. The forward and backward processes must satisfy the same reciprocity equation. Because the number of dynamical variables is usually unbalanced on both sides of a chemical equation, it is necessary that this balance be established by including as many of the dynamical variables as needed before the reciprocity equation can be formulated. Statistical transformation models of the reactions are formulated. The models are classified under the titles free exchange, restricted exchange and simplified restricted exchange. The special equations for the forward and backward processes are obtained. The models are consistent with the H theorem and Le Chatelier's principle. The models are also formulated in the context of the direct simulation Monte Carlo method.
Moral Beliefs and Cognitive Homogeneity
Directory of Open Access Journals (Sweden)
Nevia Dolcini
2018-04-01
Full Text Available The Emotional Perception Model of moral judgment intends to account for experientialism about morality and moral reasoning. In explaining how moral beliefs are formed and applied in practical reasoning, the model attempts to overcome the mismatch between reason and action/desire: morality isn’t about reason for actions, yet moral beliefs, if caused by desires, may play a motivational role in (moral agency. The account allows for two kinds of moral beliefs: genuine moral beliefs, which enjoy a relation to desire, and motivationally inert moral beliefs acquired in ways other than experience. Such etiology-based dichotomy of concepts, I will argue, leads to the undesirable view of cognition as a non-homogeneous phenomenon. Moreover, the distinction between moral beliefs and moral beliefs would entail a further dichotomy encompassing the domain of moral agency: one and the same action might possibly be either genuine moral, or not moral, if acted by individuals lacking the capacity for moral feelings, such as psychopaths.
Homogeneous modes of cosmological instantons
Energy Technology Data Exchange (ETDEWEB)
Gratton, Steven; Turok, Neil
2001-06-15
We discuss the O(4) invariant perturbation modes of cosmological instantons. These modes are spatially homogeneous in Lorentzian spacetime and thus not relevant to density perturbations. But their properties are important in establishing the meaning of the Euclidean path integral. If negative modes are present, the Euclidean path integral is not well defined, but may nevertheless be useful in an approximate description of the decay of an unstable state. When gravitational dynamics is included, counting negative modes requires a careful treatment of the conformal factor problem. We demonstrate that for an appropriate choice of coordinate on phase space, the second order Euclidean action is bounded below for normalized perturbations and has a finite number of negative modes. We prove that there is a negative mode for many gravitational instantons of the Hawking-Moss or Coleman{endash}De Luccia type, and discuss the associated spectral flow. We also investigate Hawking-Turok constrained instantons, which occur in a generic inflationary model. Implementing the regularization and constraint proposed by Kirklin, Turok and Wiseman, we find that those instantons leading to substantial inflation do not possess negative modes. Using an alternate regularization and constraint motivated by reduction from five dimensions, we find a negative mode is present. These investigations shed new light on the suitability of Euclidean quantum gravity as a potential description of our universe.
Homogeneous modes of cosmological instantons
International Nuclear Information System (INIS)
Gratton, Steven; Turok, Neil
2001-01-01
We discuss the O(4) invariant perturbation modes of cosmological instantons. These modes are spatially homogeneous in Lorentzian spacetime and thus not relevant to density perturbations. But their properties are important in establishing the meaning of the Euclidean path integral. If negative modes are present, the Euclidean path integral is not well defined, but may nevertheless be useful in an approximate description of the decay of an unstable state. When gravitational dynamics is included, counting negative modes requires a careful treatment of the conformal factor problem. We demonstrate that for an appropriate choice of coordinate on phase space, the second order Euclidean action is bounded below for normalized perturbations and has a finite number of negative modes. We prove that there is a negative mode for many gravitational instantons of the Hawking-Moss or ColemanendashDe Luccia type, and discuss the associated spectral flow. We also investigate Hawking-Turok constrained instantons, which occur in a generic inflationary model. Implementing the regularization and constraint proposed by Kirklin, Turok and Wiseman, we find that those instantons leading to substantial inflation do not possess negative modes. Using an alternate regularization and constraint motivated by reduction from five dimensions, we find a negative mode is present. These investigations shed new light on the suitability of Euclidean quantum gravity as a potential description of our universe
Nuclear reactor kinetics and control
International Nuclear Information System (INIS)
Lewins, J.
1978-01-01
A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)
Cluster-cell calculation using the method of generalized homogenization
International Nuclear Information System (INIS)
Laletin, N.I.; Boyarinov, V.F.
1988-01-01
The generalized-homogenization method (GHM), used for solving the neutron transfer equation, was applied to calculating the neutron distribution in the cluster cell with a series of cylindrical cells with cylindrically coaxial zones. Single-group calculations of the technological channel of the cell of an RBMK reactor were performed using GHM. The technological channel was understood to be the reactor channel, comprised of the zirconium rod, the water or steam-water mixture, the uranium dioxide fuel element, and the zirconium tube, together with the adjacent graphite layer. Calculations were performed for channels with no internal sources and with unit incoming current at the external boundary as well as for channels with internal sources and zero current at the external boundary. The PRAKTINETs program was used to calculate the symmetric neutron distributions in the microcell and in channels with homogenized annular zones. The ORAR-TsM program was used to calculate the antisymmetric distribution in the microcell. The accuracy of the calculations were compared for the two channel versions
International Nuclear Information System (INIS)
Merle-Lucotte, E.
2008-06-01
Within the frame of development of sustainable nuclear programs, this report does not only deal with the development of nuclear systems, but with the general context in which such a development will occur. While describing and commenting her professional career in different nuclear research institutions and on various research programs, the author describes the assets and challenges of the electro-nuclear sector, and then focuses on the research structures and contexts for future possible nuclear concepts, and more particularly like melted salt reactors for which she highlights scientific and technical problems which are still to be solved. She describes French, European and world programs which were to start by 2009
Integrability of Hamiltonian systems with homogeneous potentials of degree zero
Energy Technology Data Exchange (ETDEWEB)
Casale, Guy, E-mail: guy.casale@univ-rennes1.f [IRMAR UMR 6625, Universite de Rennes 1, Campus de Beaulieu, 35042 Rennes Cedex (France); Duval, Guillaume, E-mail: dduuvvaall@wanadoo.f [1 Chemin du Chateau, 76 430 Les Trois Pierres (France); Maciejewski, Andrzej J., E-mail: maciejka@astro.ia.uz.zgora.p [Institute of Astronomy, University of Zielona Gora, Licealna 9, PL-65-417 Zielona Gora (Poland); Przybylska, Maria, E-mail: Maria.Przybylska@astri.uni.torun.p [Torun Centre for Astronomy, N. Copernicus University, Gagarina 11, PL-87-100 Torun (Poland)
2010-01-04
We derive necessary conditions for integrability in the Liouville sense of classical Hamiltonian systems with homogeneous potentials of degree zero. We obtain these conditions through an analysis of the differential Galois group of variational equations along a particular solution generated by a non-zero solution d element of C{sup n} of nonlinear equation gradV(d)=d. We prove that when the system is integrable the Hessian matrix V{sup ''}(d) has only integer eigenvalues and is diagonalizable.
Homogeneous cosmology with aggressively expanding civilizations
International Nuclear Information System (INIS)
Jay Olson, S
2015-01-01
In the context of a homogeneous Universe, we note that the appearance of aggressively expanding advanced life is geometrically similar to the process of nucleation and bubble growth in a first-order cosmological phase transition. We exploit this similarity to describe the dynamics of life saturating the Universe on a cosmic scale, adapting the phase transition model to incorporate probability distributions of expansion and resource consumption strategies. Through a series of numerical solutions spanning several orders of magnitude in the input assumption parameters, the resulting cosmological model is used to address basic questions related to the intergalactic spreading of life, dealing with issues such as timescales, observability, competition between strategies, and first-mover advantage. Finally, we examine physical effects on the Universe itself, such as reheating and the backreaction on the evolution of the scale factor, if such life is able to control and convert a significant fraction of the available pressureless matter into radiation. We conclude that the existence of life, if certain advanced technologies are practical, could have a significant influence on the future large-scale evolution of the Universe. (paper)
Numerical computation of homogeneous slope stability.
Xiao, Shuangshuang; Li, Kemin; Ding, Xiaohua; Liu, Tong
2015-01-01
To simplify the computational process of homogeneous slope stability, improve computational accuracy, and find multiple potential slip surfaces of a complex geometric slope, this study utilized the limit equilibrium method to derive expression equations of overall and partial factors of safety. This study transformed the solution of the minimum factor of safety (FOS) to solving of a constrained nonlinear programming problem and applied an exhaustive method (EM) and particle swarm optimization algorithm (PSO) to this problem. In simple slope examples, the computational results using an EM and PSO were close to those obtained using other methods. Compared to the EM, the PSO had a small computation error and a significantly shorter computation time. As a result, the PSO could precisely calculate the slope FOS with high efficiency. The example of the multistage slope analysis indicated that this slope had two potential slip surfaces. The factors of safety were 1.1182 and 1.1560, respectively. The differences between these and the minimum FOS (1.0759) were small, but the positions of the slip surfaces were completely different than the critical slip surface (CSS).
Numerical Computation of Homogeneous Slope Stability
Directory of Open Access Journals (Sweden)
Shuangshuang Xiao
2015-01-01
Full Text Available To simplify the computational process of homogeneous slope stability, improve computational accuracy, and find multiple potential slip surfaces of a complex geometric slope, this study utilized the limit equilibrium method to derive expression equations of overall and partial factors of safety. This study transformed the solution of the minimum factor of safety (FOS to solving of a constrained nonlinear programming problem and applied an exhaustive method (EM and particle swarm optimization algorithm (PSO to this problem. In simple slope examples, the computational results using an EM and PSO were close to those obtained using other methods. Compared to the EM, the PSO had a small computation error and a significantly shorter computation time. As a result, the PSO could precisely calculate the slope FOS with high efficiency. The example of the multistage slope analysis indicated that this slope had two potential slip surfaces. The factors of safety were 1.1182 and 1.1560, respectively. The differences between these and the minimum FOS (1.0759 were small, but the positions of the slip surfaces were completely different than the critical slip surface (CSS.
Homogeneity and thermodynamic identities in geometrothermodynamics
Energy Technology Data Exchange (ETDEWEB)
Quevedo, Hernando [Universidad Nacional Autonoma de Mexico, Instituto de Ciencias Nucleares (Mexico); Universita di Roma ' ' La Sapienza' ' , Dipartimento di Fisica, Rome (Italy); ICRANet, Rome (Italy); Quevedo, Maria N. [Universidad Militar Nueva Granada, Departamento de Matematicas, Facultad de Ciencias Basicas, Bogota (Colombia); Sanchez, Alberto [CIIDET, Departamento de Posgrado, Queretaro (Mexico)
2017-03-15
We propose a classification of thermodynamic systems in terms of the homogeneity properties of their fundamental equations. Ordinary systems correspond to homogeneous functions and non-ordinary systems are given by generalized homogeneous functions. This affects the explicit form of the Gibbs-Duhem relation and Euler's identity. We show that these generalized relations can be implemented in the formalism of black hole geometrothermodynamics in order to completely fix the arbitrariness present in Legendre invariant metrics. (orig.)
A literature review on biotic homogenization
Guangmei Wang; Jingcheng Yang; Chuangdao Jiang; Hongtao Zhao; Zhidong Zhang
2009-01-01
Biotic homogenization is the process whereby the genetic, taxonomic and functional similarity of two or more biotas increases over time. As a new research agenda for conservation biogeography, biotic homogenization has become a rapidly emerging topic of interest in ecology and evolution over the past decade. However, research on this topic is rare in China. Herein, we introduce the development of the concept of biotic homogenization, and then discuss methods to quantify its three components (...
The homogeneous boundary value problem of the thick spherical shell
International Nuclear Information System (INIS)
Linder, F.
1975-01-01
With the aim to solve boundary value problems in the same manner as it is attained at thin shell theory (Superposition of Membrane solution to solution of boundary values), one has to search solutions of the equations of equilibrium of the three dimensional thick shell which produce tensions at the cut edge and are zero on the whole shell surface inside and outside. This problem was solved with the premissions of the linear theory of Elasticity. The gained solution is exact and contains the symmetric and non-symmetric behaviour and is described in relatively short analytical expressions for the deformations and tensions, after the problem of the coupled system had been solved. The static condition of the two surfaces (zero tension) leads to a homogeneous system of complex equations with the index of the Legendre spherical function as Eigenvalue. One symmetrical case is calculated numerically and is compared with the method of finite elements. This comparison results in good accordance. (Auth.)
International Nuclear Information System (INIS)
Ait Abderrahim, A.
2002-01-01
SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised
Energy Technology Data Exchange (ETDEWEB)
Ait Abderrahim, A
2001-04-01
The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.
Energy Technology Data Exchange (ETDEWEB)
Ait Abderrahim, A
2002-04-01
SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.
International Nuclear Information System (INIS)
Ait Abderrahim, A.
2001-01-01
The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised
Operating and maintenance manual for the HFIR production model homogeneity scanner
International Nuclear Information System (INIS)
Reynolds, J.W.; Shipp, R.L.; Sliski, T.F.; Longaker, W.H.; Klindt, K.K.
1984-12-01
The fuel material in a HFIR fuel is U 3 O 8 dispersed in aluminum, resembling an airfoil in cross section. To ensure uniform generation of heat within the plate, all plates must be tested (nondestructively) to determine that the U 3 O 8 content is within specified limits. The HFIR homogeneity scanner developed for this purpose is a density/thickness gauge that bombards a plate with a highly collimated, 0.062-in.-diam beam of x rays and detects those transmitted through the plate. Variations in the transmitted x rays due to absorption in the fuel plate are a measure of fuel denisty. In addition to the fuel plates for HFIR, fuel plates for several other reactors, such as the Oak Ridge Research Reactor (ORR) are also checked by the homogeneity scanner by using other sets of standards. All of the other reactors have a uniform cross section. This manual describes procedures for its electronic components
Shaw, J
2013-01-01
Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp
Russell, Charles R
1962-01-01
Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor
Present status of the use of LEU in aqueous reactors to produce Mo-99
International Nuclear Information System (INIS)
Ball, Russell M.; Pavshook, V.A.; Khvostionov, V.Ye.
1998-01-01
An operating aqueous homogeneous reactor, the ARGUS at Kurchatov Institute, has been used to produce fission product molybdenum-99 (Mo-99), widely used in nuclear medicine to produce technetium-99m (Tc-99m). The Mo-99 has been extracted from the sulfate solution using an organic sorbent after operation at 1 kW/liter. after purification, the material has been assayed and the result is well within required specification of the USPharmacopaeia. Operation calculation are presented to show the sources and quantity of alpha activity when LEU is used. (author)
Self-consolidating concrete homogeneity
Directory of Open Access Journals (Sweden)
Jarque, J. C.
2007-08-01
Full Text Available Concrete instability may lead to the non-uniform distribution of its properties. The homogeneity of self-consolidating concrete in vertically cast members was therefore explored in this study, analyzing both resistance to segregation and pore structure uniformity. To this end, two series of concretes were prepared, self-consolidating and traditional vibrated materials, with different w/c ratios and types of cement. The results showed that selfconsolidating concretes exhibit high resistance to segregation, albeit slightly lower than found in the traditional mixtures. The pore structure in the former, however, tended to be slightly more uniform, probably as a result of less intense bleeding. Such concretes are also characterized by greater bulk density, lower porosity and smaller mean pore size, which translates into a higher resistance to pressurized water. For pore diameters of over about 0.5 Î¼m, however, the pore size distribution was found to be similar to the distribution in traditional concretes, with similar absorption rates.En este trabajo se estudia la homogeneidad de los hormigones autocompactantes en piezas hormigonadas verticalmente, determinando su resistencia a la segregación y la uniformidad de su estructura porosa, dado que la pérdida de estabilidad de una mezcla puede conducir a una distribución no uniforme de sus propiedades. Para ello se han fabricado dos tipos de hormigones, uno autocompactante y otro tradicional vibrado, con diferentes relaciones a/c y distintos tipos de cemento. Los resultados ponen de manifiesto que los hormigones autocompactantes presentan una buena resistencia a la segregación, aunque algo menor que la registrada en los hormigones tradicionales. A pesar de ello, su estructura porosa tiende a ser ligeramente más uniforme, debido probablemente a un menor sangrado. Asimismo, presentan una mayor densidad aparente, una menor porosidad y un menor tamaño medio de poro, lo que les confiere mejores
CEFR information management system solution
International Nuclear Information System (INIS)
Lu Fei; Zhao Jia'ning
2011-01-01
Based on finished information resources planning scheme for China sodium cooled experimental fast breeder reactor and the advanced information resources management solution concepts were applied, we got the building solution of CEFR information management systems. At the same time, the technical solutions of systems structures, logic structures, physical structures, development platforms and operation platforms for information resources management system in fast breeder reactors were developed, which provided programmatic introductions for development works in future. (authors)
Kaplan, Renata; Erjavec, Boštjan; Senila, Marin; Pintar, Albin
2014-10-01
Catalytic wet air oxidation (CWAO) is classified as an advanced oxidation process, which proved to be highly efficient for the removal of emerging organic pollutant bisphenol A (BPA) from water. In this study, BPA was successfully removed in a batch-recycle trickle-bed reactor over bare titanate nanotube-based catalysts at very short space time of 0.6 min gCAT g(-1). The as-prepared titanate nanotubes, which underwent heat treatment at 600 °C, showed high activity for the removal of aqueous BPA. Liquid-phase recycling (5- or 10-fold recycle) enabled complete BPA conversion already at 200 °C, together with high conversion of total organic carbon (TOC), i.e., 73 and 98 %, respectively. The catalyst was chemically stable in the given range of operating conditions for 189 h on stream.
International Nuclear Information System (INIS)
Raghu, G.; Maruthi Mohan, P.; Balaji, V.; Venkateswaran, G.; Rodrigue, A.; Lyon 1 Univ., 69
2008-01-01
Removal of radioactive cobalt at trace levels (∼nM) in the presence of large excess (10 6 -fold) of corrosion product ions of complexed Fe, Cr, and Ni in spent chemical decontamination formulations (simulated effluent) of nuclear reactors is currently done by using synthetic organic ion exchangers. A large volume of solid waste is generated due to the nonspecific nature of ion sorption. Our earlier work using various fungi and bacteria, with the aim of nuclear waste volume reduction, realized up to 30% of Co removal with specific capacities calculated up to 1 μg/g in 6-24 h. In the present study using engineered Escherichia coli expressing NiCoT genes from Rhodopseudomonas palustris CGA009 (RP) and Novosphingobium aromaticivorans F-199 (NA), we report a significant increase in the specific capacity for Co removal (12 μg/g) in 1-h exposure to simulated effluent. About 85% of Co removal was achieved in a two-cycle treatment with the cloned bacteria. Expression of NiCoT genes in the E. coli knockout mutant of NiCoT efflux gene (rcnA) was more efficient as compared to expression in wild-type E. coli MC4100, JM109 and BL21 (DE3) hosts. The viability of the E. coli strains in the formulation as well as at different doses of gamma rays exposure and the effect of gamma dose on their cobalt removal capacity are determined. The potential application scheme of the above process of bioremediation of cobalt from nuclear power reactor chemical decontamination effluents is discussed. (orig.)
Investigations into homogenization of electromagnetic metamaterials
DEFF Research Database (Denmark)
Clausen, Niels Christian Jerichau
This dissertation encompasses homogenization methods, with a special interest into their applications to metamaterial homogenization. The first method studied is the Floquet-Bloch method, that is based on the assumption of a material being infinite periodic. Its field can then be expanded in term...