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)

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.

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

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

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

Measurement and regulation of the level of a homogeneous plutonium reactor; Mesure et regulation du niveau d'un reacteur homogene au plutonium

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)

Value distribution of meromorphic solutions of homogeneous and non-homogeneous complex linear differential-difference equations

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.

Neutron diffraction radiation of solid solution of carbon and hydrogen in the α-titanium in the homogeneity domain

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)

Analytical solutions of time-fractional models for homogeneous Gardner equation and non-homogeneous differential equations

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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.

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.

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)

Homogenization Theory for the Prediction of Obstructed Solute Diffusivity in Macromolecular Solutions.

Science.gov (United States)

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.

Homogenization Theory for the Prediction of Obstructed Solute Diffusivity in Macromolecular Solutions.

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.

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.

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.

Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

Science.gov (United States)

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

Homogeneous anisotropic solutions of topologically massive gravity with a cosmological constant and their homogeneous deformations

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)

Homogeneous anisotropic solutions of topologically massive gravity with a cosmological constant and their homogeneous deformations

Science.gov (United States)

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.

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

Systematic assembly homogenization and local flux reconstruction for nodal method calculations of fast reactor power distributions

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

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.

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)

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.)

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

Science.gov (United States)

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

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

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)

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

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)

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)

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

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.

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

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

Reactor of the XXI century

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 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

Feasibility neutronic conceptual design for the core configuration of a 75 kWth Aqueous Homogeneous Reactor for 99Mo production

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)

Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

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.

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)

Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel

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

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....

Homogenized blocked arcs for multicriteria optimization of radiotherapy: Analytical and numerical solutions

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

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

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)

Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

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.

Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

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

Large Marks-decahedral Pd nanoparticles synthesized by a modified hydrothermal method using a homogeneous reactor

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.

Large Marks-decahedral Pd nanoparticles synthesized by a modified hydrothermal method using a homogeneous reactor

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.

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

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)

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.)

Homogeneous solutions of stationary Navier-Stokes equations with isolated singularities on the unit sphere. II. Classification of axisymmetric no-swirl solutions

Science.gov (United States)

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.

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.)

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

Interpretation of criticality experiments on homogeneous solutions of plutonium and uranium; Interpretation des experiences de criticite sur des solutions homogenes de plutonium et d'uranium

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)

Fast reactors

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.)

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

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.

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

Study of the boron homogenizing process employing an experimental low-pressure bench simulating the IRIS reactor pressurizer – Part I

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

Study of the boron homogenizing process employing an experimental low-pressure bench simulating the IRIS reactor pressurizer – Part II

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

Single-particle levitation system for automated study of homogeneous solute nucleation

OpenAIRE

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...

Homogeneous Solutions of Stationary Navier-Stokes Equations with Isolated Singularities on the Unit Sphere. I. One Singularity

Science.gov (United States)

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.

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.)

Analytic solutions of the multigroup space-time reactor kinetics equations

International Nuclear Information System (INIS)

Lee, C.E.; Rottler, S.

1986-01-01

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

An 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)

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

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)

On the modeling of irradiation-induced homogeneous precipitation in proton-bombarded Ni-Si solid solutions

Science.gov (United States)

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

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

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.

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

Travelling wave solutions of the homogeneous one-dimensional FREFLO model

Science.gov (United States)

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.

Contribution to analytical solution of neutron slowing down problem in homogeneous and heterogeneous media

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

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)

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.

The analytical benchmark solution of spatial diffusion kinetics in source driven systems for homogeneous media

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)

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

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)

The application of polynomial chaos methods to a point kinetics model of MIPR: An Aqueous Homogeneous Reactor

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

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)

Degradation of aqueous phenol solutions by coaxial DBD reactor

Science.gov (United States)

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).

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)

The treatment of absorber rod heterogeneity effects using homogeneous equivalent cross-sections and their application in large fast reactors

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

Optimization study and neutronic and thermal-hydraulic design calculations of a 75 KWTH aqueous homogeneous reactor for medical isotopes production

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)

Degradation of nicotine in water solutions using a water falling film DBD plasma reactor: direct and indirect treatment

Science.gov (United States)

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.

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

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.

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

Homogenization of the coefficient of diffusion: influence of modelling and of the laplacian for fast power reactors and experimental mockups

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

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)

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.

Homogenization of neutronic diffusion models; Homogeneisation des modeles de diffusion en neutronique

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)

Cost optimization of ADS design: Comparative study of externally driven heterogeneous and homogeneous two-zone subcritical reactor systems

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

Computer modelling of fuel behaviour and performance in research reactors as related to fuel depletion

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)

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

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.

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

Chemical state analysis of iron(III) compounds precipitated homogeneously from solutions containing urea by means of Moessbauer spectrometry and x-ray diffractometry

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)

Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions

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.

Explicit finite-difference solution of two-dimensional solute transport with periodic flow in homogenous porous media

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.

Heterogeneous reactors

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

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)

Synthesis and characterization of homogeneous interstitial solutions of nitrogen and carbon in iron-based lattices

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...

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)

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.)

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.

Nuclear power technology system with molten salt reactor for transuranium nuclides burning in closed fuel cycle

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)

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

Analytical representation of the solution of the space kinetic diffusion equation in a one-dimensional and homogeneous domain

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)

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)

Numerical reactor evaluation

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

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

A model to estimate volume change due to radiolytic gas bubbles and thermal expansion in solution reactors

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)

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

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

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.

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

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)

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

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.

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

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

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

Electro-magnetostatic homogenization of bianisotropic metamaterials

OpenAIRE

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...

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

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)

Numerical method for solution of transient, homogeneous, equilibrium, two-phase flows in one space dimension

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

Solution XAS Analysis for Exploring the Active Species in Homogeneous Vanadium Complex Catalysis

Science.gov (United States)

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.

LPV model development and control of a solution copolymerization reactor

NARCIS (Netherlands)

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

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

Design of a homogeneous subcritical nuclear reactor based on thorium with a source of californium 252; Diseno de un reactor nuclear subcritico homogeneo a base de Torio con una fuente de Californio 252

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)

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)

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.

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

Analytical Solution for Time-drawdown Response to Constant Pumping from a Homogeneous, Confined Horizontal Aquifer with Unidirectional Flow

Science.gov (United States)

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

Highly Dispersed Pseudo-Homogeneous and Heterogeneous Catalysts Synthesized via Inverse Micelle Solutions for the Liquefaction of Coal

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.

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)

Multilevel Monte Carlo Approaches for Numerical Homogenization

KAUST Repository

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.

Homogeneous solutions of hydrophilic enzymes in nonpolar organic solvents. New systems for fundamental studies and biocatalytic transformations.

Science.gov (United States)

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).

Fluid-structure-interaction of the pressurized water reactor core internals during blowdown - numerical simulation with a homogenization model

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

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.

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)

Analytical solution of spatial kinetics of the diffusion model for subcritical homogeneous systems driven by external source

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)

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

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.

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

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.

Flows and chemical reactions in homogeneous mixtures

CERN Document Server

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

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)

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)

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.

Solutes transport in unsaturated double-porosity medium. Modelling by homogenization and applications

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)

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

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)

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

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)

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.

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)

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

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)

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)

Multilevel Monte Carlo Approaches for Numerical Homogenization

KAUST Repository

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

LSENS: A General Chemical Kinetics and Sensitivity Analysis Code for homogeneous gas-phase reactions. Part 1: Theory and numerical solution procedures

Science.gov (United States)

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.

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.)

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

International Nuclear Information System (INIS)

Park, Jong Woon

2009-01-01

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

Continuous energy Monte Carlo method based homogenization multi-group constants calculation

International Nuclear Information System (INIS)

Li Mancang; Wang Kan; Yao Dong

2012-01-01

The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)

Continuous formation of N-chloro-N,N-dialkylamine solutions in well-mixed meso-scale flow reactors

Science.gov (United States)

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.

Novel electrode structure in a DBD reactor applied to the degradation of phenol in aqueous solution

Science.gov (United States)

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.

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

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1996-07-01

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

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

Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores; Homogeneisation de modeles de transferts thermiques et radiatifs: application au coeur des reacteurs a caloporteur gaz

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)

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)

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

Complete set of homogeneous isotropic analytic solutions in scalar-tensor cosmology with radiation and curvature

Science.gov (United States)

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.

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.)

Computation of the homogeneous and forced solutions of a finite length, line-driven, submerged plate.

Science.gov (United States)

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.

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)

Probing structural homogeneity of La{sub 1-x}Gd{sub x}PO{sub 4} monazite-type solid solutions by combined spectroscopic and computational studies

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.

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.

Benchmarking homogenization algorithms for monthly data

Science.gov (United States)

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.

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

[Computer modeling the hydrostatic pressure characteristics of the membrane potential for polymeric membrane, separated non-homogeneous electrolyte solutions].

Science.gov (United States)

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.

Contribution to analytical solution of neutron slowing down problem in homogeneous and heterogeneous media; Prilog analitickom resavanju problema usporavanja neutrona u homogenim i heterogenim sredinama

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.

Flexible solution of linear program with an application to decommissioning planning of nuclear reactor

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.)

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

An efficient technique for the point reactor kinetics equations with Newtonian temperature feedback effects

International Nuclear Information System (INIS)

Nahla, Abdallah A.

2011-01-01

Highlights: → An efficient technique for the nonlinear reactor kinetics equations is presented. → This method is based on Backward Euler or Crank Nicholson and fundamental matrix. → Stability of efficient technique is defined and discussed. → This method is applied to point kinetics equations of six-groups of delayed neutrons. → Step, ramp, sinusoidal and temperature feedback reactivities are discussed. - Abstract: The point reactor kinetics equations of multi-group of delayed neutrons in the presence Newtonian temperature feedback effects are a system of stiff nonlinear ordinary differential equations which have not any exact analytical solution. The efficient technique for this nonlinear system is based on changing this nonlinear system to a linear system by the predicted value of reactivity and solving this linear system using the fundamental matrix of the homogenous linear differential equations. The nonlinear point reactor kinetics equations are rewritten in the matrix form. The solution of this matrix form is introduced. This solution contains the exponential function of a variable coefficient matrix. This coefficient matrix contains the unknown variable, reactivity. The predicted values of reactivity in the explicit form are determined replacing the exponential function of the coefficient matrix by two kinds, Backward Euler and Crank Nicholson, of the rational approximations. The nonlinear point kinetics equations changed to a linear system of the homogenous differential equations. The fundamental matrix of this linear system is calculated using the eigenvalues and the corresponding eigenvectors of the coefficient matrix. Stability of the efficient technique is defined and discussed. The efficient technique is applied to the point kinetics equations of six-groups of delayed neutrons with step, ramp, sinusoidal and the temperature feedback reactivities. The results of these efficient techniques are compared with the traditional methods.

Homogenization for rigid suspensions with random velocity-dependent interfacial forces

KAUST Repository

Gorb, Yuliya

2014-12-01

We study suspensions of solid particles in a viscous incompressible fluid in the presence of random velocity-dependent interfacial forces. The flow at a small Reynolds number is modeled by the Stokes equations, coupled with the motion of rigid particles arranged in a periodic array. The objective is to perform homogenization for the given suspension and obtain an equivalent description of a homogeneous (effective) medium, the macroscopic effect of the interfacial forces and the effective viscosity are determined using the analysis on a periodicity cell. In particular, the solutions uωε to a family of problems corresponding to the size of microstructure ε and describing suspensions of rigid particles with random surface forces imposed on the interface, converge H1-weakly as ε→0 a.s. to a solution of a Stokes homogenized problem, with velocity dependent body forces. A corrector to a homogenized solution that yields a strong H1-convergence is also determined. The main technical construction is built upon the Γ-convergence theory. © 2014 Elsevier Inc.

An analytic solution to the homogeneous EIT problem on the 2D disk and its application to estimation of electrode contact impedances

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

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)

Layout optimization using the homogenization method

Science.gov (United States)

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.

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.

Intracortical stiffness of mid-diaphysis femur bovine bone: lacunar-canalicular based homogenization numerical solutions and microhardness measurements.

Science.gov (United States)

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.

Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

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

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

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)

Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

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

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

Modification of a Turbulent Boundary Layer within a Homogeneous Concentration of Drag reducing Polymer Solution

Science.gov (United States)

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 Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

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.

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

Benchmarking monthly homogenization algorithms

Science.gov (United States)

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

Fuel solution criticality accident studies with the SILENE reactor: phenomenology, consequences and simulated intervention

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

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

Fast Spectrum Reactors

CERN Document Server

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...

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

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

Improvement of the homogeneity of atomized particles dispersed in high uranium density research reactor fuels

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)

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.)

Feasibility study of a solar reactor for phenol treatment by the Photo-Fenton process in aqueous solution

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)

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

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.)

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)

Forward flux sampling calculation of homogeneous nucleation rates from aqueous NaCl solutions.

Science.gov (United States)

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.

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

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)

Magnetite synthesis from ferrous iron solution at pH 6.8 in a continuous stirred tank reactor.

Science.gov (United States)

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.

JAERI thermal reactor standard code system for reactor design and analysis SRAC

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1985-01-01

SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

Techniques to limit gaseous releases in case of reactor accident. Choice criteria - present solutions

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

Shielding computations for solution transfer lines from Analytical Lab to process cells of Demonstration Fast Reactor Plant (DFRP)

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

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.)

Monte Carlo simulation of neutron transport in a homogeneous reactor with a partially inserted control rod

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

Methods to homogenize electrochemical concentration cell (ECC ozonesonde measurements across changes in sensing solution concentration or ozonesonde manufacturer

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 �

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

Two-dimensional Haar wavelet Collocation Method for the solution of Stationary Neutron Transport Equation in a homogeneous isotropic medium

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

Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling

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.)

Fourier-Accelerated Nodal Solvers (FANS) for homogenization problems

Science.gov (United States)

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.

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)

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

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

Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

Science.gov (United States)

Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

2017-07-01

In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

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.

Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

International Nuclear Information System (INIS)

Siefman, Daniel J.; Girardin, Gaëtan; Rais, Adolfo; Pautz, Andreas; Hursin, Mathieu

2015-01-01

Highlights: • Modeling of research reactors. • Serpent and PARCS coupling. • Lattice physics codes modeling techniques. - Abstract: This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of k eff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of k eff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients

Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

International Nuclear Information System (INIS)

Kocik, J.; Keilova, E.

1993-01-01

A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

1994-12-31

A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

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)

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.

A homogenization method for the analysis of a coupled fluid-structure interaction problem with inner solid structure

International Nuclear Information System (INIS)

Sigrist, Jean-Francois; Laine, Christian; Broc, Daniel

2006-01-01

The present paper exposes a homogenization method developed in order to perform the seismic analysis of a nuclear reactor with internal structures modelling and taking fluid structure interaction effects into account. The numerical resolution of fluid-structure interactions has made tremendous progress over the past decades and some applications of the various developed techniques in the industrial field can be found in the literature. As builder of nuclear naval propulsion reactors (ground prototype reactor or embarked reactor on submarines), DCN Propulsion has been working with French nuclear committee CEA for several years in order to integrate fluid-structure analysis in the design stage of current projects. In previous papers modal and seismic analyses of a nuclear reactor with fluid-structure interaction effect were exposed. The studies highlighted the importance of fluid- structure coupling phenomena in the industrial case and focussed on added mass and added stiffness effects. The numerical model used in the previous studies did not take into account the presence of internal structures within the pressure vessel. The present study aims at improving the numerical model of the nuclear reactor to take into account the presence of the internal structures. As the internal structures are periodical within the inner and outer structure of the pressure vessel the proposed model is based on the development of a homogenization method: the presence of internal structure and its effect on the fluid-structure physical interaction is taken into account, although they are not geometrically modeled. The basic theory of the proposed homogenization method is recalled, leading to the modification of fluid-structure coupling operator in the finite element model. The physical consistency of the method is proved by an evaluation of the system mass with the various mass operators (structure, fluid and fluid-structure operators). The method is exposed and validated in a 2 D case

Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

International Nuclear Information System (INIS)

Lee, K.B.; Im, I.Y.; Lee, S.K.

1992-01-01

An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

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)

[Computer modeling the dependences of the membrane potential for polymeric membrane separated non-homogeneous electrolyte solutions on concentration Rayleigh number].

Science.gov (United States)

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).

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)

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)

Multiregion reactors

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

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

A Singular Perturbation Problem for Steady State Conversion of Methane Oxidation in Reverse Flow Reactor

Directory of Open Access Journals (Sweden)

Aang Nuryaman

2012-11-01

Full Text Available The governing equations describing the methane oxidation process in reverse flow reactor are given by a set of convective-diffusion equations with a nonlinear reaction term, where temperature and methane conversion are dependent variables. In this study, the process is assumed to be one-dimensional pseudo homogeneous model and takes place with a certain reaction rate in which the whole process of reactor is still workable. Thus, the reaction rate can proceed at a fixed temperature. Under this condition, we restrict ourselves to solve the equations for the conversion only. From the available data, it turns out that the ratio of the diffusion term to the reaction term is small. Hence, this ratio is considered as small parameter in our model and this leads to a singular perturbation problem. In the vicinity of small parameter in front of higher order term, the numerical difficulties will be found. Here, we present an analytical solution by means of matched asymptotic expansions. Result shows that, up to and including the first order of approximation, the solution is in agreement with the exact and numerical solutions of the boundary value problem.

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

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Nuclear reactor

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

CONIFERS: a neutronics code for reactors with channels

International Nuclear Information System (INIS)

Davis, R.S.

1977-04-01

CONIFERS is a neutronics code for nuclear reactors whose fuel is in channels that are separated from each other by several neutron mean-free-path lengths of moderator. It can treat accurately situations in which the usual homogenized-cell diffusion equation becomes inaccurate, but is more economical than other advanced methods such as response-matrix and source-sink formalisms. CONIFERS uses exact solutions of the neutron diffusion equation within each cell. It allows for the breakdown of this equation near a channel by means of data that almost any cell code can supply. It uses the results of these cell analyses in a reactor equations set that is as readily solvable as the familiar finite-difference equations set. CONIFERS can model almost any configuration of channels and other structures in two or three dimensions. It can use any number of energy groups and any reactivity scales, including scales based on control operations. It is also flexible from a programming point of view, and has convenient input and output provisions. (author)

Solution of Point Reactor Neutron Kinetics Equations with Temperature Feedback by Singularly Perturbed Method

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.

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

Homogenization of Large-Scale Movement Models in Ecology

Science.gov (United States)

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.

Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects

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)

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.

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

A safeguards verification technique for solution homogeneity and volume measurements in process tanks

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

Engineering solutions for a reflector change concept in the high-temperature reactor with pebble bed core and OTTO-fueling

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

Systematic homogenization and self-consistent flux and pin power reconstruction for nodal diffusion methods. 1: Diffusion equation-based theory

International Nuclear Information System (INIS)

Zhang, H.; Rizwan-uddin; Dorning, J.J.

1995-01-01

A diffusion equation-based systematic homogenization theory and a self-consistent dehomogenization theory for fuel assemblies have been developed for use with coarse-mesh nodal diffusion calculations of light water reactors. The theoretical development is based on a multiple-scales asymptotic expansion carried out through second order in a small parameter, the ratio of the average diffusion length to the reactor characteristic dimension. By starting from the neutron diffusion equation for a three-dimensional heterogeneous medium and introducing two spatial scales, the development systematically yields an assembly-homogenized global diffusion equation with self-consistent expressions for the assembly-homogenized diffusion tensor elements and cross sections and assembly-surface-flux discontinuity factors. The rector eigenvalue 1/k eff is shown to be obtained to the second order in the small parameter, and the heterogeneous diffusion theory flux is shown to be obtained to leading order in that parameter. The latter of these two results provides a natural procedure for the reconstruction of the local fluxes and the determination of pin powers, even though homogenized assemblies are used in the global nodal diffusion calculation

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

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

Response matrix method for neutron transport in reactor lattices using group symmetry properties

International Nuclear Information System (INIS)

Mund, E.H.

1991-01-01

This paper describes a response matrix method for the approximate solution of one-velocity, multi-dimensional transport problems in reactor lattices, with isotropic neutron scattering. The transport equation is solved on a homogeneous cell by using a Petrov-Galerkin technique based on a set of trial and test functions (including polynomials and exponential functions) closely related to transport problems in infinite media. The number of non-zero elements of the response matrices reduces to a minimum when the symmetry properties of the cell are included ab initio in the span of the basis functions. To include these properties, use is made of projection operations which are performed very efficiently on symbolic manipulation programs. Numerical results of model problems in square geometry show a good agreement with reference solutions

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)

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

A scheme to calculate higher-order homogenization as applied to micro-acoustic boundary value problems

Science.gov (United States)

Vagh, Hardik A.; Baghai-Wadji, Alireza

2008-12-01

Current technological challenges in materials science and high-tech device industry require the solution of boundary value problems (BVPs) involving regions of various scales, e.g. multiple thin layers, fibre-reinforced composites, and nano/micro pores. In most cases straightforward application of standard variational techniques to BVPs of practical relevance necessarily leads to unsatisfactorily ill-conditioned analytical and/or numerical results. To remedy the computational challenges associated with sub-sectional heterogeneities various sophisticated homogenization techniques need to be employed. Homogenization refers to the systematic process of smoothing out the sub-structural heterogeneities, leading to the determination of effective constitutive coefficients. Ordinarily, homogenization involves a sophisticated averaging and asymptotic order analysis to obtain solutions. In the majority of the cases only zero-order terms are constructed due to the complexity of the processes involved. In this paper we propose a constructive scheme for obtaining homogenized solutions involving higher order terms, and thus, guaranteeing higher accuracy and greater robustness of the numerical results. We present

Kinetic Study of COS with Tertiary Alkanolamine Solutions. 2. Modeling and Experiments in a Stirred Cell Reactor

NARCIS (Netherlands)

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

WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

International Nuclear Information System (INIS)

2004-01-01

1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

Neutron thermalization in absorbing infinite homogeneous media: theoretical methods; Methodes theoriques pour l'etude de la thermalisation des neutrons dans les milieux absorbants infinis et homogenes

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)

Interim Report on Fluid-Fuel Thermal Breeder Reactors (Revised)

International Nuclear Information System (INIS)

MacPherson, H. G.; Alexander, L. G.; Carter, W. L.; Chapman, R. H.; Kinyon, B. W.; Miller, J. W.

1960-01-01

The merits of aqueous-homogeneous ), graphite-moderated molten salt (MSBR) , and graphite-moderated liquid-bismuth (LBBR) breeder reactors operated at nearly comparable fuel-cycle costs (~1.5 mills/kwhr) were evaluated. The net electrical plant capability was assumed to be 1000 MwE, and the fuel and fertile streams were processed continuously on-site. The specific powers based on fuel were 1.2, 1.2, and 0.5 MwE/kg respectively, and 5.9, 3.7, and 5.3 MwE/tonne based on thorium. Net breeding ratios were 1.10, 1.07, and 1.07, giving doubling times of 5-1/2, 11, and 25 full power years . The fuel-cycle costs at the design points selected were 1.4, 1.3, and 1.6 mills/kwhr . The AHBR has an advantage in breeding ratio and doubling time because D 2 O is superior to graphite as a moderator in breeder reactors. MSBR has an advantage in fuel-cycle costs and in inventory of uranium in the fertile stream as a result of using a solution blanket.

A study of the homogeneous stages in the catalytic oxidation of naphthalene, o-xylene, and benzene over a vibratory-fluidized catalyst bed

Energy Technology Data Exchange (ETDEWEB)

Korneichuk, G P; Stasevich, V P; Shaprinskaya, T M; Girushtin, G G; Gritsenko, V I; Zelenchukova, T G

1978-01-01

To identify the conditions for minimizing homogeneous states, the reaction kinetics were studied in a vibrating gradientless quartz reactor both in the presence and absence of the catalyst. A tenfold decrease of the reactional space in the absence of catalyst inhibited the oxidation (e.g., from a 68% conversion to 10% at 500/sup 0/C for o-xylene, and from 100% to 2% at 580/sup 0/C for benzene), whereas increasing the surface-volume ratio of the reactor increased the oxidation rate for benzene, which indicated that noncatalytic oxidation follows a radical-chain mechanism and involves both homogeneous (mainly) and heterogeneous stages. Catalytic oxidation carried out in a small volume (to avoid the homogeneous states) followed a heterogeneous mechanism up to 580/sup 0/C for naphthalene and o-xylene, and up to 550/sup 0/C for benzene. At higher temperatures, however, volume oxidation of benzene to carbon oxides was detected, which was favored by intense reactor vibration (i.e., increasing free space between catalyst grains), constituted 27% at 564/sup 0/C and 40% at 584/sup 0/C, and probably followed a heterogeneous-homogeneous mechanism. The partial oxidation products (i.e., phthalic and (for benzene) maleic anhydride) formed entirely by a heterogeneous mechanism. Tables and graphs.

Collision-free gases in spatially homogeneous space-times

International Nuclear Information System (INIS)

Maartens, R.; Maharaj, S.D.

1985-01-01

The kinematical and dynamical properties of one-component collision-free gases in spatially homogeneous, locally rotationally symmetric (LRS) space-times are analyzed. Following Ray and Zimmerman [Nuovo Cimento B 42, 183 (1977)], it is assumed that the distribution function f of the gas inherits the symmetry of space-time, in order to construct solutions of Liouville's equation. The redundancy of their further assumption that f be based on Killing vector constants of the motion is shown. The Ray and Zimmerman results for Kantowski--Sachs space-time are extended to all spatially homogeneous LRS space-times. It is shown that in all these space-times the kinematic average four-velocity u/sup i/ can be tilted relative to the homogeneous hypersurfaces. This differs from the perfect fluid case, in which only one space-time admits tilted u/sup i/, as shown by King and Ellis [Commun. Math. Phys. 31, 209 (1973)]. As a consequence, it is shown that all space-times admit nonzero acceleration and heat flow, while a subclass admits nonzero vorticity. The stress π/sub i/j is proportional to the shear sigma/sub i/j by virtue of the invariance of the distribution function. The evolution of tilt and the existence of perfect fluid solutions is also discussed

Nuclear reactors theory

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

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.)

Study of the fluctuations in neutrons density in a homogeneous plutonium pile; Etude de la fluctuation de la population de neutrons sur une pile homogene au plutonium

Energy Technology Data Exchange (ETDEWEB)

Clouet d' Orval, C; Deilgat, E; Labbe, J; Molbert, M; Tachon, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1962-07-01

The variations in neutron density in the centre of a reactor core are subjected to a statistical fluctuation phenomenon. Because of the correlations existing between the neutrons, their distribution does not follow Poisson's law. It diverges from it by a quantity depending on various parameter such as the lifetime and the effective fraction of the slowed-down neutrons. Experiments have been carried out designed to demonstrate this divergence and also, more particularly, the correlation between unaffected and slowed-down neutrons; these have been carried out with the help of the reactor 'Proserpine', a critical homogeneous plutonium assembly. (author) [French] L'evolution de la population neutronique au sein d'un milieu multiplicateur est soumise a un phenomene de fluctuations statistiques. En raison des correlations qui existent entre les neutrons, leur distribution ne suit pas une loi de Poisson. Elle s'en ecarte d'une quantite qui depend de divers parametres tels que le temps de vie, la proportion effective des neutrons retardes. Des experiences, destinees a mettre en evidence cet ecart, et plus particulierement les correlations entre neutrons prompts et neutrons retardes, ont ete realisees a l'aide de la pile 'Proserpine', experience critique homogene au plutonium. (auteur)

Response of thermal multi zone reactors to local perturbation of reactivity; Odziv termalnih multizonih reaktora na lokalnu perturbaciju reaktivnosti

Energy Technology Data Exchange (ETDEWEB)

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

1967-01-15

A modal analysis method, spatial eigenfunctions expansion, was applied for solving the time dependent diffusion equation in two-group approximation. The absorption cross section in the thermal neutron range was time dependent. The response functions for radially multi-zone reactor systems and homogeneous reactors were obtained as solutions of the time dependent diffusion equation. Some numerical results obtained by this method are included. U radu je primenjena modalna analiza, razvoj po prostornim svojstvenim funkcijama, za resavanje vremenski zavisne difuzione jednacine u dvogrupnoj aproksimaciji, kada je presek za apsorpciju u termalnoj grupi funkcija vremena. Kao rezultat resavanja vremenski zavisne difuzione jednacine dobijeni su izrazi za prenosne funkcije radijalno multizonih reaktorskih sistema i homogenih reaktorskih sistema. Dati su i neki numericki rezultati primene ove metode (author)

Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations

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.

Uranium-plutonium fuel for fast reactors

International Nuclear Information System (INIS)

Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

1996-01-01

Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

Small Modular Reactors for Countries with Small to Medium Electric Grids - An Economically Sensible Solution

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

Calculation of homogenized Pickering NGS stainless steel adjuster rod neutron cross-sections using conservation of reaction rates

Energy Technology Data Exchange (ETDEWEB)

Robinson, R C [Atlantic Nuclear Services Ltd. (Canada); Tran, F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

A homogenization methodology for calculation of reactivity device incremental cross-sections has been developed using reaction rate conservation (RRC). A heterogeneous transport calculation of flux was utilised to produce the homogenized cross-sections for a finite difference two group diffusion code. The RRC cross-sections have been shown to improve significantly the prediction of reactivity worth for stainless steel adjuster rods installed in Pickering NGS reactors. (author). 10 refs., 3 tabs., 6 figs.

DRAGON, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the

Homogeneity of nuclear fuel containing burnable poison; Homogenost jedrskega goriva z gorljivim strupom

Energy Technology Data Exchange (ETDEWEB)

Loose, A; Susnik, D; Ilic, R [Institut Jozef Stefan, Ljubljana (Yugoslavia)

1988-07-01

In this work the results of the microstructural investigations of the influence of the Gd{sub 2}O{sub 3} contents and the sintering conditions on the formation of the homogeneous (U,Gd)O{sub 2} solid solution, are presented. For this purpose sintering conditions, microstructure and diffusivity in UO{sub 2} -Gd{sub 2}O{sub 3} , were studied. It was found that, with a suitable preparation of powders and longer sintering times in dry hydrogen atmosphere above 1700 deg C, a homogeneous (U,Gd)O{sub 2} solid solution can be obtained. (author)

Features of the Numerical Solution of Thermal Destruction Fuel Pins Problems in the Fast Reactor

Science.gov (United States)

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.

The theoretical possibility of reducing the doubling time in a fast-reactor by using heterogeneous configurations of various types of fuel

International Nuclear Information System (INIS)

Orlov, V.V.; Slesarev, I.S.; Zaritskij, S.M.; Subbotin, S.A.; Alekseev, P.N.; Zverkov, Yu.A.

1980-01-01

The authors have derived approximate expressions relating the doubling time of a fast reactor using various types of fuel simultaneously to the doubling time of traditional (homogeneous) reactors in which these types of fuel are used separately. These relationships afford a means of determining the conditions in which the use of various types of fuel can result in an improved doubling time. It was established that the use of heterogeneous compositions formed from assemblies of homogeneous systems gives a notable gain in doubling time over that of any of the original homogeneous systems if the doubling times were similar to each other. This gain is fairly large even in the case of BN reactors with high fuel volume fractions. The size of the gain depends on the degree of ''differentiation'' in the neutron and thermal properties of the components of the heterogeneous reactor. An optimum proportion has been found for the assemblies taken from the original homogeneous systems, governed primarily by the ratio of fuel densities. Estimates were made of the advantages of metallic oxide compositions over the traditional compositions used in large, fast reactors of the BN type. These estimates indicate that the former can be considered as alternative homogeneous compositions with carbide or nitride fuel as far as breeding characteristics are concerned. (author)

A numerical technique for enhanced efficiency and stability for the solution of the nuclear reactor equation

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)

A numerical technique for enhanced efficiency and stability for the solution of the nuclear reactor equation

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)

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)

Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO32 aqueous solution reactor driven by a 30-MeV proton accelerator

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.

Nuclear-Thermal Analysis of Fully Ceramic Microencapsulated Fuel via Two-Temperature Homogenized Model

International Nuclear Information System (INIS)

Lee, Yoon Hee; Cho, Nam Zin

2013-01-01

The FCM fuel is based on a proven safety philosophy that has been utilized operationally in very high temperature reactors (VHTRs). However, the FCM fuel consists of TRISO particles randomly dispersed in SiC matrix. The high heterogeneity in composition leads to difficulty in explicit thermal calculation of such a fuel. Therefore, an appropriate homogenization model becomes essential. In this paper, we apply the two-temperature homogenized model to thermal analysis of an FCM fuel. The model was recently proposed in order to provide more realistic temperature profiles in the fuel element in VHTRs. We applied the two-temperature homogenized model to FCM fuel. The two-temperature homogenized model was obtained by particle transport Monte Carlo calculation applied to the pellet region consisting of many coated particles uniformly dispersed in SiC matrix. Since this model gives realistic temperature profiles in the pellet (providing fuel-kernel temperature and SiC matrix temperature distinctly), it can be used for more accurate neutronics evaluation such as Doppler temperature feedback. The transient thermal calculation may be performed also more realistically with temperature-dependent homogenized parameters in various scenarios

Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor

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.)

Novel Budesonide Particles for Dry Powder Inhalation Prepared Using a Microfluidic Reactor Coupled With Ultrasonic Spray Freeze Drying.

Science.gov (United States)

Saboti, Denis; Maver, Uroš; Chan, Hak-Kim; Planinšek, Odon

2017-07-01

Budesonide (BDS) is a potent active pharmaceutical ingredient, often administered using respiratory devices such as metered dose inhalers, nebulizers, and dry powder inhalers. Inhalable drug particles are conventionally produced by crystallization followed by milling. This approach tends to generate partially amorphous materials that require post-processing to improve the formulations' stability. Other methods involve homogenization or precipitation and often require the use of stabilizers, mostly surfactants. The purpose of this study was therefore to develop a novel method for preparation of fine BDS particles using a microfluidic reactor coupled with ultrasonic spray freeze drying, and hence avoiding the need of additional homogenization or stabilizer use. A T-junction microfluidic reactor was employed to produce particle suspension (using an ethanol-water, methanol-water, and an acetone-water system), which was directly fed into an ultrasonic atomization probe, followed by direct feeding to liquid nitrogen. Freeze drying was the final preparation step. The result was fine crystalline BDS powders which, when blended with lactose and dispersed in an Aerolizer at 100 L/min, generated fine particle fraction in the range 47.6% ± 2.8% to 54.9% ± 1.8%, thus exhibiting a good aerosol performance. Subsequent sample analysis confirmed the suitability of the developed method to produce inhalable drug particles without additional homogenization or stabilizers. The developed method provides a viable solution for particle isolation in microfluidics in general. Copyright © 2017 American Pharmacists Association®. All rights reserved.

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

The performance of 3500 MWth homogeneous and heterogeneous metal fueled core designs

International Nuclear Information System (INIS)

Turski, R.; Yang, Shi-tien.

1987-11-01

Performance parameters are calculated for a representative 3500 MWth homogeneous and a heterogeneous metal fueled reactor design. The equilibrium cycle neutronic characteristics, safety coefficients, control system requirements, and control rod worths are evaluated. The thermal-hydraulic characteristics for both configurations are also compared. The heavy metal fuel loading requirements and neutronic performance characteristics are also evaluated for the uranium startup option. 14 refs., 14 figs., 20 tabs

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)

Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

International Nuclear Information System (INIS)

Blumenfeld, P.E.

1995-08-01

Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

Dynamic Behavior of Reverse Flow Reactor for Lean Methane Combustion

OpenAIRE

Yogi W. Budhi; M. Effendy; Yazid Bindar; Subagjo

2014-01-01

The stability of reactor operation for catalytic oxidation of lean CH4 has been investigated through modeling and simulation, particularly the influence of switching time and heat extraction on reverse flow reactor (RFR) performance. A mathematical model of the RFR was developed, based on one-dimensional pseudo-homogeneous model for mass and heat balances, incorporating heat loss through the reactor wall. The configuration of the RFR consisted of inert-catalyst-inert, with or without heat ext...

Testing Homogeneity with the Galaxy Fossil Record

CERN Document Server

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 (0.2homogeneity, we calculate the posterior distribution for the excess large-scale variance due to inhomogeneity, and find that the most likely solution is n...

A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

International Nuclear Information System (INIS)

Limaiem, I.

2006-12-01

Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

Integrated, Reactor Relevant Solutions for Lower Hybrid Range of Frequencies Actuators

Science.gov (United States)

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.

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

Mechanical Homogenization Increases Bacterial Homogeneity in Sputum

Science.gov (United States)

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

Use of heterogeneous finite elements generated by collision probability solutions to calculate a pool reactor core

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

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)

MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors

International Nuclear Information System (INIS)

Williams, M.M.R.

2007-01-01

Description: Prof. M.M..R Williams has now released three of his legacy books for free distribution: 1 - M.M.R. Williams: The Slowing Down and Thermalization of Neutrons, North-Holland Publishing Company - Amsterdam, 582 pages, 1966. Content: Part I - The Thermal Energy Region: 1. Introduction and Historical Review, 2. The Scattering Kernel, 3. Neutron Thermalization in an Infinite Homogeneous Medium, 4. Neutron Thermalization in Finite Media, 5. The Spatial Dependence of the Energy Spectrum, 6. Reactor Cell Calculations, 7. Synthetic Scattering Kernels. Part II - The Slowing Down Region: 8. Scattering Kernels in the Slowing Down Region, 9. Neutron Slowing Down in an Infinite Homogeneous Medium, 10.Neutron Slowing Down and Diffusion. 2 - M.M.R. Williams: Mathematical Methods in Particle Transport Theory, Butterworths, London, 430 pages, 1971. Content: 1 The General Problem of Particle Transport, 2 The Boltzmann Equation for Gas Atoms and Neutrons, 3 Boundary Conditions, 4 Scattering Kernels, 5 Some Basic Problems in Neutron Transport and Rarefied Gas Dynamics, 6 The Integral Form of the Transport Equation in Plane, Spherical and Cylindrical Geometries, 7 Exact Solutions of Model Problems, 8 Eigenvalue Problems in Transport Theory, 9 Collision Probability Methods, 10 Variational Methods, 11 Polynomial Approximations. 3 - M.M.R. Williams: Random Processes in Nuclear Reactors, Pergamon Press Oxford New York Toronto Sydney, 243 pages, 1974. Content: 1. Historical Survey and General Discussion, 2. Introductory Mathematical Treatment, 3. Applications of the General Theory, 4. Practical Applications of the Probability Distribution, 5. The Langevin Technique, 6. Point Model Power Reactor Noise, 7. The Spatial Variation of Reactor Noise, 8. Random Phenomena in Heterogeneous Reactor Systems, 9. Associated Fluctuation Problems, Appendix: Noise Equivalent Sources. Note to the user: Prof. M.M.R Williams owns the copyright of these books and he authorises the OECD/NEA Data Bank

Reactor kinetics methods development. Final report

International Nuclear Information System (INIS)

Hansen, K.F.; Henry, A.F.

1978-01-01

This report is a qualitative summary of research conducted at MIT from 1967 to 1977 in the area of reactor kinetics methods. The objectives of the research were to find methods of integration of various mathematical models of nuclear reactor transients. From the beginning the work was aimed at numerical integration methods. Specific areas of research, discussed in more detail following, included: integration of multigroup diffusion theory models by finite difference and finite element methods; response matrix and nodal methods; coarse-mesh homogenization; and special treatment of boundary conditions

Exact traveling wave solutions of the Boussinesq equation

International Nuclear Information System (INIS)

Ding Shuangshuang; Zhao Xiqiang

2006-01-01

The repeated homogeneous balance method is used to construct exact traveling wave solutions of the Boussinesq equation, in which the homogeneous balance method is applied to solve the Riccati equation and the reduced nonlinear ordinary differential equation, respectively. Many new exact traveling wave solutions of the Boussinesq equation are successfully obtained

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)

Solidification Segregation and Homogenization Behavior of 1Cr-1.25Mo-0.25V Steel Ingot

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Bae [Dae-gu Mechatronics and Materials Institute, Daegu (Korea, Republic of); Na, Young-Sang; Seo, Seong-Moon [Korea Institute of Materials Science, Changwon (Korea, Republic of); Lee, Je-Hyun [Changwon National University, Changwon (Korea, Republic of)

2016-09-15

As a first step to optimizing the homogenization heat treatment following high temperature upset forging, the solidification segregation and the homogenization behaviors of solute elements were quantitatively analyzed for 1Cr-1.25Mo-0.25V steel ingot by electron probe micro-analysis (EPMA). The random sampling approach, which was designed to generate continuous compositional profiles of each solute element, was employed to clarify the segregation and homogenization behaviors. In addition, ingot castings of lab-scale and a 16-ton-sized 1Cr-1.25Mo-0.25V steel were simulated using the finite element method in three dimensions to understand the size effect of the ingot on the microsegregation and its reduction during the homogenization heat treatment. It was found that the microsegregation in a large-sized ingot was significantly reduced by the promotion of solid state diffusion due to the extremely low cooling rate. On the other hand, from the homogenization point of view, increasing the ingot size causes a dramatic increase in the dendrite arm spacing, and hence the homogenization of microsegregation in a large-sized ingot appears to be practically difficult.

SERPENT Monte Carlo reactor physics code

International Nuclear Information System (INIS)

Leppaenen, J.

2010-01-01

SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

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)

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

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)

Simulating the Behaviour of the Fast Reactor Joyo (Draft)

International Nuclear Information System (INIS)

Juutilainen, Pauli

2008-01-01

Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO 2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

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

Application of non-thermal plasma reactor and Fenton reaction for degradation of ibuprofen

Energy Technology Data Exchange (ETDEWEB)

Marković, Marijana [Center of Chemistry, Institute of Chemistry, Technology and Metallurgy, University of Belgrade, Studentski trg 12-16, 11000 Belgrade (Serbia); Jović, Milica; Stanković, Dalibor [Innovation Center, Faculty of Chemistry, University of Belgrade, P.O. Box 51, 11058 Belgrade 118 (Serbia); Kovačević, Vesna [Faculty of Physics, University of Belgrade, P.O. Box 44, 11000 Belgrade (Serbia); Roglić, Goran [Faculty of Chemistry, University of Belgrade, P.O. Box 51, 11058 Belgrade 118 (Serbia); Gojgić-Cvijović, Gordana [Center of Chemistry, Institute of Chemistry, Technology and Metallurgy, University of Belgrade, Studentski trg 12-16, 11000 Belgrade (Serbia); Manojlović, Dragan, E-mail: manojlo@chem.bg.ac.rs [Faculty of Chemistry, University of Belgrade, P.O. Box 51, 11058 Belgrade 118 (Serbia)

2015-02-01

Pharmaceutical compounds have been detected frequently in surface and ground water. Advanced Oxidation Processes (AOPs) were reported as very efficient for removal of various organic compounds. Nevertheless, due to incomplete degradation, toxic intermediates can induce more severe effects than the parent compound. Therefore, toxicity studies are necessary for the evaluation of possible uses of AOPs. In this study the effectiveness and capacity for environmental application of three different AOPs were estimated. They were applied and evaluated for removal of ibuprofen from water solutions. Therefore, two treatments were performed in a non-thermal plasma reactor with dielectric barrier discharge with and without a homogenous catalyst (Fe{sup 2+}). The third treatment was the Fenton reaction. The degradation rate of ibuprofen was measured by HPLC-DAD and the main degradation products were identified using LC–MS TOF. Twelve degradation products were identified, and there were differences according to the various treatments applied. Toxicity effects were determined with two bioassays: Vibrio fischeri and Artemia salina. The efficiency of AOPs was demonstrated for all treatments, where after 15 min degradation percentage was over 80% accompanied by opening of the aromatic ring. In the treatment with homogenous catalyst degradation reached 99%. V. fischeri toxicity test has shown greater sensitivity to ibuprofen solution after the Fenton treatment in comparison to A. salina. - Highlights: • Twelve ibuprofen degradation products were identified in total. • The degradation percentage differed between treatments (DBD/Fe{sup 2+} was 99%). • In DBD/Fe{sup 2+} only aliphatic degradation products were identified. • V. fischeri was sensitive to ibuprofen solution after the Fenton treatment. • A. salina showed no toxic effect when exposed to all post treatment solutions.

Application of non-thermal plasma reactor and Fenton reaction for degradation of ibuprofen

International Nuclear Information System (INIS)

Marković, Marijana; Jović, Milica; Stanković, Dalibor; Kovačević, Vesna; Roglić, Goran; Gojgić-Cvijović, Gordana; Manojlović, Dragan

2015-01-01

Pharmaceutical compounds have been detected frequently in surface and ground water. Advanced Oxidation Processes (AOPs) were reported as very efficient for removal of various organic compounds. Nevertheless, due to incomplete degradation, toxic intermediates can induce more severe effects than the parent compound. Therefore, toxicity studies are necessary for the evaluation of possible uses of AOPs. In this study the effectiveness and capacity for environmental application of three different AOPs were estimated. They were applied and evaluated for removal of ibuprofen from water solutions. Therefore, two treatments were performed in a non-thermal plasma reactor with dielectric barrier discharge with and without a homogenous catalyst (Fe 2+ ). The third treatment was the Fenton reaction. The degradation rate of ibuprofen was measured by HPLC-DAD and the main degradation products were identified using LC–MS TOF. Twelve degradation products were identified, and there were differences according to the various treatments applied. Toxicity effects were determined with two bioassays: Vibrio fischeri and Artemia salina. The efficiency of AOPs was demonstrated for all treatments, where after 15 min degradation percentage was over 80% accompanied by opening of the aromatic ring. In the treatment with homogenous catalyst degradation reached 99%. V. fischeri toxicity test has shown greater sensitivity to ibuprofen solution after the Fenton treatment in comparison to A. salina. - Highlights: • Twelve ibuprofen degradation products were identified in total. • The degradation percentage differed between treatments (DBD/Fe 2+ was 99%). • In DBD/Fe 2+ only aliphatic degradation products were identified. • V. fischeri was sensitive to ibuprofen solution after the Fenton treatment. • A. salina showed no toxic effect when exposed to all post treatment solutions

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

An elastoplastic homogenization procedure for predicting the settlement of a foundation on a soil reinforced by columns

OpenAIRE

ABDELKRIM, Malek; DE BUHAN, Patrick

2007-01-01

This paper presents an elastoplastic homogenization method applied to a soil reinforced by regularly distributed columns. According to this method, the composite reinforced soil is regarded, from a macroscopic point of view, as a homogeneous anisotropic continuous medium, the elastic as well as plastic properties of which can be obtained from the solution to an auxiliary problem attached to the reinforced soil representative cell. Based upon an approximate solution to this problem, in which p...

WATER BOILER REACTOR

Science.gov (United States)

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

Controlling hydrophilicity of polymer film by altering gas flow rate in atmospheric-pressure homogeneous plasma

International Nuclear Information System (INIS)

Kang, Woo Seok; Hur, Min; Lee, Jae-Ok; Song, Young-Hoon

2014-01-01

Graphical abstract: - Highlights: • Controlling hydrophilicity of polymer film by varying gas flow rate is proposed in atmospheric-pressure homogeneous plasma treatment. • Without employing additional reactive gas, requiring more plasma power and longer treatment time, hydrophilicity of polyimide films was improved after the low-gas-flow plasma treatment. • The gas flow rate affects the hydrophilic properties of polymer surface by changing the discharge atmosphere in the particular geometry of the reactor developed. • Low-gas-flow induced wettability control suggests effective and economical plasma treatment. - Abstract: This paper reports on controlling the hydrophilicity of polyimide films using atmospheric-pressure homogeneous plasmas by changing only the gas flow rate. The gas flow changed the discharge atmosphere by mixing the feed gas with ambient air because of the particular geometry of the reactor developed for the study, and a low gas flow rate was found to be favorable because it generated abundant nitrogen or oxygen species that served as sources of hydrophilic functional groups over the polymer surface. After low-gas-flow plasma treatment, the polymer surface exhibited hydrophilic characteristics with increased surface roughness and enhanced chemical properties owing to the surface addition of functional groups. Without adding any reactive gases or requiring high plasma power and longer treatment time, the developed reactor with low-gas-flow operation offered effective and economical wettability control of polyimide films

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

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

ADVANCED CONTROL FOR A ETHYLENE REACTOR

Directory of Open Access Journals (Sweden)

Dumitru POPESCU

2017-06-01

Full Text Available The main objective of this work is the design and implementation of control solutions for petrochemical processes, namely the control and optimization of a pyrolysis reactor, the key-installation in the petrochemical industry. We present the technological characteristics of this petrochemical process and some aspects about the proposed control system solution for the ethylene plant. Finally, an optimal operating point for the reactor is found, considering that the process has a nonlinear multi-variable structure. The results have been implemented on an assembly of pyrolysis reactors on a petrochemical platform from Romania.

Device for the generation of homogeneous dose distributions in irradiated materials

International Nuclear Information System (INIS)

Leonhardt, J.; Schulze, H.; Boes, J.; Decker, U.; Schmidt, J.

1985-01-01

The invention has been directed at a device for the generation of homogeneous dose distributions in materials irradiated by charged particles. This device can be applied to the initiation of radiation-chemical reactions in solids, of cross-linking and vulcanizing reactors, of crystal defect annealings, etc. A movable absorber (e.g. a wedge or a solid of revolution) which periodically changes the energy of particles striking the specimen has been installed in the beam hole of the beam generating system

NUCLEAR REACTOR

Science.gov (United States)

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.

Directions in advanced reactor technology

International Nuclear Information System (INIS)

Golay, M.W.

1990-01-01

Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state

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)

FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix

International Nuclear Information System (INIS)

Misfeldt, I.B.

1987-01-01

1 - Nature of physical problem solved: The two-dimensional neutron diffusion equation (xy geometry) is solved in the homogeneous form (K eff calculation). The boundary conditions specify each group current as a linear homogeneous function of the group fluxes (gamma matrix concept). For each material, the fission matrix is assumed to be dyadic. 2 - Method of solution: Finite element formulation with Lagrange type elements. Solution technique: SOR with extrapolation. 3 - Restrictions on the complexity of the problem: Maximum order of the Lagrange elements is 6

Study of an hypothetical reactor meltdown accident for a 50 MW sub(th) fast reactor

International Nuclear Information System (INIS)

Azevedo, E.M. de.

1983-01-01

A melhodology for determining the energy released in hypothetical reactor meltdown accidents is presented. A numerical code was developed based upon the Nicholson method for a uniform and homogeneous reactor with spherical geometry. A comparative study with other know programs in the literature which use better approximations for small energy released, shows that the methodology used were compatible with those under comparison. Besides the influence of some parameters on the energy released, such as the initial power level and the prompt neutron lifetime was studied under this metodology and its result exhibitted. The Doppler effect was also analyzed and its influence on the energy released has been emphasized. (Author) [pt

Electrical model of dielectric barrier discharge homogenous and filamentary modes

Science.gov (United States)

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.

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)

Electro-recovery of gold and silver from a cyanide leaching solution using a three-dimensional reactor

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.

Electro-recovery of gold and silver from a cyanide leaching solution using a three-dimensional reactor

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

Actinide recycling in reactors; Aktiniden-Rezyklierung in Reaktoren

Energy Technology Data Exchange (ETDEWEB)

Kuesters, H.; Wiese, H.W.; Krieg, B.

1995-08-01

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)

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.)

Economic Issues of Fast Reactor in China

International Nuclear Information System (INIS)

Yang Hongyi

2013-01-01

Conclusions: 1. More and more fast reactors could be appearing in the world currently and near future. 2. China gets little experience and practice about the economics issues of sodium cooled fast reactors. 3. The economic issues become more and more important for the deplot of fast reactors. Suggestions: 1. An authoritative economic evaluation solution for fast reactor and related fuel cycles facilities is necessary. The solution may be developed by the interested country in order to share the few data, experience and methodology. 2. A new initiative to help to share the economic information for fast reactor and related fuel cycle facilities is necessary. A meeting like TM-44899 organized by the IAEA is very beneficial for this topic and hopefully will continue

A computational analysis on homogeneous-heterogeneous mechanism in Carreau fluid flow

Science.gov (United States)

Khan, Imad; Rehman, Khalil Ur; Malik, M. Y.; Shafquatullah

2018-03-01

In this article magnetohydrodynamic Carreau fluid flow towards stretching cylinder is considered in the presence of homogeneous-heterogeneous reactions effect. The flow model is structured by utilizing theoretical grounds. For the numerical solution a shooting method along with Runge-Kutta algorithm is executed. The outcomes are provided through graphs. It is observed that the Carreau fluid concentration shows decline values via positive iterations of homogeneous-heterogeneous reaction parameters towards both shear thinning and thickening case. The present work is certified through comparison with already existing literature in a limiting sense.

Optimal truss and frame design from projected homogenization-based topology optimization

DEFF Research Database (Denmark)

Larsen, S. D.; Sigmund, O.; Groen, J. P.

2018-01-01

In this article, we propose a novel method to obtain a near-optimal frame structure, based on the solution of a homogenization-based topology optimization model. The presented approach exploits the equivalence between Michell’s problem of least-weight trusses and a compliance minimization problem...... using optimal rank-2 laminates in the low volume fraction limit. In a fully automated procedure, a discrete structure is extracted from the homogenization-based continuum model. This near-optimal structure is post-optimized as a frame, where the bending stiffness is continuously decreased, to allow...

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

1 kWe sodium borohydride hydrogen generation system Part II: Reactor modeling

OpenAIRE

Zhang, Jinsong; Zheng, Yuan; Gore, Jay P; Mudawar, Issam; Fisher, Timothy

2007-01-01

Sodium borohydride (NaBH4) hydrogen storage systems offer many advantages for hydrogen storage applications. The physical processes inside a NaBH4 packed bed reactor involve multi-component and multi-phase flow and multi-mode heat and mass transfer. These processes are also coupled with reaction kinetics. To guide reactor design and optimization, a reactor model involving all of these processes is desired. A onedimensional numerical model in conjunction with the assumption of homogeneous cata...

Oscillation estimates relative to p-homogeneous forms and Kato measures data

Directory of Open Access Journals (Sweden)

Marco Biroli

2006-11-01

Full Text Available We state pointwise estimate for the positive subsolutions associated to a p-homogeneous form and nonnegative Radon measures data. As a by-product we establish an oscillation’s estimate for the solutions relative to Kato measures data.

The epithermal neutron-flux distribution in the reactor RA - Vinca

International Nuclear Information System (INIS)

Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

1987-01-01

The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

Vacuum solutions of Bianchi cosmologies in quadratic gravity

International Nuclear Information System (INIS)

Deus, Juliano Alves de; Muller, Daniel

2011-01-01

Full text: In this work we solve numerically the vacuum solutions of field equations of Bianchi homogeneous universes in the context of Semiclassical theory. Our interest is to study the quadratic theory of gravity with regard in the cosmological description of our universe in periods of intense fields. Bianchi cosmologies are anisotropic homogeneous cosmological models, but can include the isotropic models as particular cases (Bianchi I, VII and IX include homogeneous and isotropic Friedmann models plane, hyperbolic and spherical, respectively). Homogeneous models are good cosmological representations of our universe. With focus in solutions for intense fields, like the early universe, where isotropy is not necessarily required, the adopted scenario is the vacuum solutions, where the geometry is dominant in determining the gravitation. Still following in this way, the Semiclassical theory, which considers quantum matter fields propagating in classical geometrical background, is addressed to give the field equations. This formalism leads to fourth-order ordinary differential equations, in contrast to second-order equations from General Relativity. The Lagrangian of the theory is quadratic in the Ricci scalar and in the Ricci tensor. The equations system is highly non-linear and can be only numerically solved, except perhaps for few particular cases. We obtained numerical solutions for Bianchi V II A evolving to Minkowski and to de Sitter solutions, and also to singularities. The both first and second solutions were obtained choosing initial conditions near from respective exact vacuum solutions from Einstein theory, which are also exact solutions of the quadratic theory. Other Bianchi types are still under study. (author)

Analysis of dryout behaviour in laterally non-homogeneous debris beds using the MEWA-2D code

International Nuclear Information System (INIS)

Rahman, Saidur; Buerger, Manfred; Buck, Michael; Pohlner, Georg; Kulenovic, Rudi; Nayak, Arun Kumar; Sehgal, Bal Raj

2009-01-01

The present study analyses the impact of lateral non-homogeneities on the coolability of heated, initially water filled debris beds. Debris beds which may be formed in a postulated severe accident in light water reactors can not be expected to have a homogeneous structure. Lateral non-homogeneities are given e.g. already by a variation in height as in a heap of debris. Internally, less porous or more porous region may occur, the latter even as downcomer-like structures are considered to favour supply of water to the bed and thus coolability. In previous work it has been shown that such non-homogeneities are often strongly enhancing coolability, as compared to earlier investigations on laterally homogeneous beds. The present contribution aims at extending the view by analysing further cases of non-homogeneities with the MEWA-2D code. Especially, effects of capillary forces are considered in contrast to earlier analysis. Part of the paper deals with specific experiments performed in the POMECO facility at KTH in which a laterally stratified debris bed has been considered, whereby especially a strong jump of porosity, from 0.26 to 0.38, has been established. Astonishingly, under top as well as bottom flooding, dryout in these experiments occurred first in the lateral layer with higher porosity. Understanding is now provided by the effect of capillary forces: water is drawn from this layer to the less porous one. This effect improves the cooling in the less porous layer while it reduces coolability of the more porous layer. No real loop behaviour of inflow via the higher porosities with subsequent upflow in the less porous layer establishes here, in contrast to expectations. Other cases (different lateral heating in an otherwise homogeneous bed, closed downcomer in a homogeneous bed and heap-like debris) show, on the other hand, strongly improved coolability by such loops establishing due to the lateral differences in void and the corresponding pressure differences

TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry

International Nuclear Information System (INIS)

Kristiansen, G.K.

1987-01-01

1 - Nature of physical problem solved: The two-dimensional neutron diffusion equation (x-y, r-z, or r-theta geometry is solved, either in the inhomogeneous (source calculation) or the homogeneous form (K eff calculation or absorber adjustment). The boundary conditions specify each group current as a linear homogeneous function of the group fluxes (gamma matrix concept). For each material, the fission matrix is assumed to by dyadic. 2 - Method of solution: Finite difference formulation (5 point scheme, mesh corner variant) is used. Solution technique: multi-line SOR. Eigenvalue estimate by neutron balance

Multigrid method applied to the solution of an elliptic, generalized eigenvalue problem

Energy Technology Data Exchange (ETDEWEB)

Alchalabi, R.M. [BOC Group, Murray Hill, NJ (United States); Turinsky, P.J. [North Carolina State Univ., Raleigh, NC (United States)

1996-12-31

The work presented in this paper is concerned with the development of an efficient MG algorithm for the solution of an elliptic, generalized eigenvalue problem. The application is specifically applied to the multigroup neutron diffusion equation which is discretized by utilizing the Nodal Expansion Method (NEM). The underlying relaxation method is the Power Method, also known as the (Outer-Inner Method). The inner iterations are completed using Multi-color Line SOR, and the outer iterations are accelerated using Chebyshev Semi-iterative Method. Furthermore, the MG algorithm utilizes the consistent homogenization concept to construct the restriction operator, and a form function as a prolongation operator. The MG algorithm was integrated into the reactor neutronic analysis code NESTLE, and numerical results were obtained from solving production type benchmark problems.

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.

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.

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

Reaction-transport simulations of non-oxidative methane conversion with continuous hydrogen removal: Homogeneous-heterogeneous methane reaction pathways

International Nuclear Information System (INIS)

Li, Lin; Borry, Richard W.; Iglesia, Enrique

2000-01-01

Detailed kinetic-transport models were used to explore thermodynamic and kinetic barriers in the non-oxidative conversion of CH4 via homogeneous and homogeneous-heterogeneous pathways and the effects of continuous hydrogen removal and of catalytic sites on attainable yields of useful C2-C10 products. The homogeneous kinetic model combines separately developed models for low-conversion pyrolysis and for chain growth to form large aromatics and carbon. The H2 formed in the reaction decreases CH4 pyrolysis rates and equilibrium conversions and it favors the formation of lighter products. The removal of H2 along tubular reactors with permeable walls increases reaction rates and equilibrium CH4 conversions. C2-C10 yields reach values greater than 90 percent at intermediate values of dimensionless transport rates (delta=1-10), defined as the ratio hydrogen transport and methane conversion rates. Homogeneous reactions require impractical residence times, even with H2 removal, because of slow initiation and chain transfer rates. The introduction of heterogeneous chain initiation pathways using surface sites that form methyl radicals eliminates the induction period without influencing the homogeneous product distribution. Methane conversion, however, occurs predominately in the chain transfer regime, within which individual transfer steps and the formation of C2 intermediates become limited by thermodynamic constraints. Catalytic sites alone cannot overcome these constraints. Catalytic membrane reactors with continuous H2 removal remove these thermodynamic obstacles and decrease the required residence time. Reaction rates become limited by homogeneous reactions of C2 products to form C6+ aromatics. Higher delta values lead to subsequent conversion of the desired C2-C10 products to larger polynuclear aromatics. We conclude that catalytic methane pyrolysis at the low temperatures required for restricted chain growth and the elimination of thermodynamics constraints via

Homogenization versus homogenization-free method to measure muscle glycogen fractions.

Science.gov (United States)

Mojibi, N; Rasouli, M

2016-12-01

The glycogen is extracted from animal tissues with or without homogenization using cold perchloric acid. Three methods were compared for determination of glycogen in rat muscle at different physiological states. Two groups of five rats were kept at rest or 45 minutes muscular activity. The glycogen fractions were extracted and measured by using three methods. The data of homogenization method shows that total glycogen decreased following 45 min physical activity and the change occurred entirely in acid soluble glycogen (ASG), while AIG did not change significantly. Similar results were obtained by using "total-glycogen-fractionation methods". The findings of "homogenization-free method" indicate that the acid insoluble fraction (AIG) was the main portion of muscle glycogen and the majority of changes occurred in AIG fraction. The results of "homogenization method" are identical with "total glycogen fractionation", but differ with "homogenization-free" protocol. The ASG fraction is the major portion of muscle glycogen and is more metabolically active form.

Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

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)

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

Effect of dynamic high pressure homogenization on the aggregation state of soy protein.

Science.gov (United States)

Keerati-U-Rai, Maneephan; Corredig, Milena

2009-05-13

Although soy proteins are often employed as functional ingredients in oil-water emulsions, very little is known about the aggregation state of the proteins in solution and whether any changes occur to soy protein dispersions during homogenization. The effect of dynamic high pressure homogenization on the aggregation state of the proteins was investigated using microdifferential scanning calorimetry and high performance size exclusion chromatography coupled with multiangle laser light scattering. Soy protein isolates as well as glycinin and beta-conglycinin fractions were prepared from defatted soy flakes and redispersed in 50 mM sodium phosphate buffer at pH 7.4. The dispersions were then subjected to homogenization at two different pressures, 26 and 65 MPa. The results demonstrated that dynamic high pressure homogenization causes changes in the supramolecular structure of the soy proteins. Both beta-conglycinin and glycinin samples had an increased temperature of denaturation after homogenization. The chromatographic elution profile showed a reduction in the aggregate concentration with homogenization pressure for beta-conglycinin and an increase in the size of the soluble aggregates for glycinin and soy protein isolate.

Homogenizing Advanced Alloys: Thermodynamic and Kinetic Simulations Followed by Experimental Results

Science.gov (United States)

Jablonski, Paul D.; Hawk, Jeffrey A.

2017-01-01

Segregation of solute elements occurs in nearly all metal alloys during solidification. The resultant elemental partitioning can severely degrade as-cast material properties and lead to difficulties during post-processing (e.g., hot shorts and incipient melting). Many cast articles are subjected to a homogenization heat treatment in order to minimize segregation and improve their performance. Traditionally, homogenization heat treatments are based upon past practice or time-consuming trial and error experiments. Through the use of thermodynamic and kinetic modeling software, NETL has designed a systematic method to optimize homogenization heat treatments. Use of the method allows engineers and researchers to homogenize casting chemistries to levels appropriate for a given application. The method also allows for the adjustment of heat treatment schedules to fit limitations on in-house equipment (capability, reliability, etc.) while maintaining clear numeric targets for segregation reduction. In this approach, the Scheil module within Thermo-Calc is used to predict the as-cast segregation present within an alloy, and then diffusion controlled transformations is used to model homogenization kinetics as a function of time and temperature. Examples of computationally designed heat treatments and verification of their effects on segregation and properties of real castings are presented.

Calculated investigation of actinide transmutation in the BOR-60 reactor

International Nuclear Information System (INIS)

Zhemkov, I.Yu.; Ishunina, O.V.; Yakovleva, I.V.

2000-01-01

One of the prospective actinide burner reactor type is the fast reactor with a 'hard' spectrum and small breeding factor, which is the BOR-60. The calculated investigations demonstrate that Loading up to 40% of minor-actinides to the BOR-60 reactor did not lead to the considerable change of neutron-physical characteristics. The performed calculations show that the BOR- 60 reactor possesses a high efficiency of the minor-actinide and plutonium bum-up (up to 37 kg/(TW · h)) hat is comparable with properties of the actinide burner-reactors under design. The BOR-60 reactor can provide a homogeneous minor-actinide Loading (minor-actinide addition to the standard fuel) to the core and heterogeneous Loading (as separate assemblies-targets with a high minor-actinide fraction) to the first rows of a radial blanket that allows the optimum usage of the reactor and its characteristics. (authors)

Nuclear reactor with a fixed system of neutron poison, which can be burnt up, introduced into the reactor core

International Nuclear Information System (INIS)

Mueller, E.; Roegler, H.J.; Wickert, M.

1985-01-01

The fixed system consists of neutron poison which can be burnt up, in an uneven distribution, and with adjustable absorber rods for output control, which are driven into the reactor core from the side along the fuel elements. There is an excess of neutron poison which can be burnt up, overall, on the side of the reactor core away from the absorber rods. The reactor core is free of neutron poison which can be burnt up on the side where the absorber rods are driven in, so that the ratio of maximum to mean power density with reference to a possible absorber rod positions is less than for homogeneous distribution of the neutron poison which can be burnt up. (orig./HP) [de

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.)

Development of a Practical Hydrogen Storage System Based on Liquid Organic Hydrogen Carriers and a Homogeneous Catalyst

Energy Technology Data Exchange (ETDEWEB)

Jensen, Craig [Hawaii Hydrogen Carriers, LLC, Honolulu, HI (United States); Brayton, Daniel [Hawaii Hydrogen Carriers, LLC, Honolulu, HI (United States); Jorgensen, Scott W. [General Motors, LLC, Warren, MI (United States). Research and Development Center. Chemical and Material Systems Lab.; Hou, Peter [General Motors, LLC, Warren, MI (United States). Research and Development Center. Chemical and Material Systems Lab.

2017-03-24

The objectives of this project were: 1) optimize a hydrogen storage media based on LOC/homogeneous pincer catalyst (carried out at Hawaii Hydrogen Carriers, LLC) and 2) develop space, mass and energy efficient tank and reactor system to house and release hydrogen from the media (carried out at General Motor Research Center).

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. 

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

International Nuclear Information System (INIS)

Ougouag, A.M.; Ferrer, R.M.

2010-01-01

The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these 'spread' the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

Method of producing homogeneous mixed metal oxides and metal--metal oxide mixtures

International Nuclear Information System (INIS)

Quinby, T.C.

1978-01-01

Metal powders, metal oxide powders, and mixtures thereof of controlled particle size are provided by reacting an aqueous solution containing dissolved metal values with excess urea. Upon heating, urea reacts with water from the solution to leave a molten urea solution containing the metal values. The molten urea solution is heated to above about 180 0 C, whereupon metal values precipitate homogeneously as a powder. The powder is reduced to metal or calcined to form oxide particles. One or more metal oxides in a mixture can be selectively reduced to produce metal particles or a mixture of metal and metal oxide particles

Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

Directory of Open Access Journals (Sweden)

Sayer C.

2004-01-01

Full Text Available This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homogeneous composition.

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (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 =.

Fibonacci-like Differential Equations with a Polynomial Non-Homogeneous Part

NARCIS (Netherlands)

Asveld, P.R.J.

1989-01-01

We investigate non-homogeneous linear differential equations of the form $x''(t) + x'(t) - x(t) = p(t)$ where $p(t)$ is either a polynomial or a factorial polynomial in $t$. We express the solution of these differential equations in terms of the coefficients of $p(t)$, in the initial conditions, and

Homogenization of one-dimensional draining through heterogeneous porous media including higher-order approximations

Science.gov (United States)

Anderson, Daniel M.; McLaughlin, Richard M.; Miller, Cass T.

2018-02-01

We examine a mathematical model of one-dimensional draining of a fluid through a periodically-layered porous medium. A porous medium, initially saturated with a fluid of a high density is assumed to drain out the bottom of the porous medium with a second lighter fluid replacing the draining fluid. We assume that the draining layer is sufficiently dense that the dynamics of the lighter fluid can be neglected with respect to the dynamics of the heavier draining fluid and that the height of the draining fluid, represented as a free boundary in the model, evolves in time. In this context, we neglect interfacial tension effects at the boundary between the two fluids. We show that this problem admits an exact solution. Our primary objective is to develop a homogenization theory in which we find not only leading-order, or effective, trends but also capture higher-order corrections to these effective draining rates. The approximate solution obtained by this homogenization theory is compared to the exact solution for two cases: (1) the permeability of the porous medium varies smoothly but rapidly and (2) the permeability varies as a piecewise constant function representing discrete layers of alternating high/low permeability. In both cases we are able to show that the corrections in the homogenization theory accurately predict the position of the free boundary moving through the porous medium.

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.

Transmutation of nuclear waste in nuclear reactors

International Nuclear Information System (INIS)

Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

1996-03-01

The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

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

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

International Nuclear Information System (INIS)

Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

1987-01-01

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

Removal of NO from flue gas by aqueous chlorine-dioxide scrubbing solution in a lab-scale bubbling reactor.

Science.gov (United States)

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.

Homogenous UV/periodate process in treatment of p-nitrophenol aqueous solutions under mild operating conditions.

Science.gov (United States)

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.

From analytical solutions of solute transport equations to multidimensional time-domain random walk (TDRW) algorithms

Science.gov (United States)

Bodin, Jacques

2015-03-01

In this study, new multi-dimensional time-domain random walk (TDRW) algorithms are derived from approximate one-dimensional (1-D), two-dimensional (2-D), and three-dimensional (3-D) analytical solutions of the advection-dispersion equation and from exact 1-D, 2-D, and 3-D analytical solutions of the pure-diffusion equation. These algorithms enable the calculation of both the time required for a particle to travel a specified distance in a homogeneous medium and the mass recovery at the observation point, which may be incomplete due to 2-D or 3-D transverse dispersion or diffusion. The method is extended to heterogeneous media, represented as a piecewise collection of homogeneous media. The particle motion is then decomposed along a series of intermediate checkpoints located on the medium interface boundaries. The accuracy of the multi-dimensional TDRW method is verified against (i) exact analytical solutions of solute transport in homogeneous media and (ii) finite-difference simulations in a synthetic 2-D heterogeneous medium of simple geometry. The results demonstrate that the method is ideally suited to purely diffusive transport and to advection-dispersion transport problems dominated by advection. Conversely, the method is not recommended for highly dispersive transport problems because the accuracy of the advection-dispersion TDRW algorithms degrades rapidly for a low Péclet number, consistent with the accuracy limit of the approximate analytical solutions. The proposed approach provides a unified methodology for deriving multi-dimensional time-domain particle equations and may be applicable to other mathematical transport models, provided that appropriate analytical solutions are available.

Homogenization of locally resonant acoustic metamaterials towards an emergent enriched continuum.

Science.gov (United States)

Sridhar, A; Kouznetsova, V G; Geers, M G D

This contribution presents a novel homogenization technique for modeling heterogeneous materials with micro-inertia effects such as locally resonant acoustic metamaterials. Linear elastodynamics is used to model the micro and macro scale problems and an extended first order Computational Homogenization framework is used to establish the coupling. Craig Bampton Mode Synthesis is then applied to solve and eliminate the microscale problem, resulting in a compact closed form description of the microdynamics that accurately captures the Local Resonance phenomena. The resulting equations represent an enriched continuum in which additional kinematic degrees of freedom emerge to account for Local Resonance effects which would otherwise be absent in a classical continuum. Such an approach retains the accuracy and robustness offered by a standard Computational Homogenization implementation, whereby the problem and the computational time are reduced to the on-line solution of one scale only.

Stochastic Stokes' Drift, Homogenized Functional Inequalities, and Large Time Behavior of Brownian Ratchets

KAUST Repository

Blanchet, Adrien

2009-01-01

A periodic perturbation of a Gaussian measure modifies the sharp constants in Poincarae and logarithmic Sobolev inequalities in the homogeniz ation limit, that is, when the period of a periodic perturbation converges to zero. We use variational techniques to determine the homogenized constants and get optimal convergence rates toward s equilibrium of the solutions of the perturbed diffusion equations. The study of these sharp constants is motivated by the study of the stochastic Stokes\\' drift. It also applies to Brownian ratchets and molecular motors in biology. We first establish a transport phenomenon. Asymptotically, the center of mass of the solution moves with a constant velocity, which is determined by a doubly periodic problem. In the reference frame attached to the center of mass, the behavior of the solution is governed at large scale by a diffusion with a modified diffusion coefficient. Using the homogenized logarithmic Sobolev inequality, we prove that the solution converges in self-similar variables attached to t he center of mass to a stationary solution of a Fokker-Planck equation modulated by a periodic perturbation with fast oscillations, with an explicit rate. We also give an asymptotic expansion of the traveling diffusion front corresponding to the stochastic Stokes\\' drift with given potential flow. © 2009 Society for Industrial and Applied Mathematics.

7 CFR 58.920 - Homogenization.

Science.gov (United States)

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...

A spatially homogeneous and isotropic Einstein-Dirac cosmology

Science.gov (United States)

Finster, Felix; Hainzl, Christian

2011-04-01

We consider a spatially homogeneous and isotropic cosmological model where Dirac spinors are coupled to classical gravity. For the Dirac spinors we choose a Hartree-Fock ansatz where all one-particle wave functions are coherent and have the same momentum. If the scale function is large, the universe behaves like the classical Friedmann dust solution. If however the scale function is small, quantum effects lead to oscillations of the energy-momentum tensor. It is shown numerically and proven analytically that these quantum oscillations can prevent the formation of a big bang or big crunch singularity. The energy conditions are analyzed. We prove the existence of time-periodic solutions which go through an infinite number of expansion and contraction cycles.

Homogeneous Reactor Experiment (HRE) Pond cryogenic barrier technology demonstration: Pre-barrier subsurface hydrology and contaminant transport investigation

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

Reactor kinetics - pulse and steady state

Energy Technology Data Exchange (ETDEWEB)

Estes, B F; Morris, F M [Sandia Laboratories (United States)

1974-07-01

An analytical model has been developed which couples the nuclear and thermal characteristics of the Annular Core Pulse Reactor (ACPR) into a solution which describes both the neutron kinetics of the reactor and the temperature behavior of a fuel-moderator element. The model describes both pulse and steady state operations. This paper describes the important aspects of the reactor, the fuel- moderator elements, the neutron kinetic equations of the reactor, and the time-temperature behavior of a fuel-moderator element that is being subjected to the maximum power density in the core. The parameters which are utilized in the equations are divided into two classes, those that can be measured directly and those that are assumed to be known (each is described briefly). Some of the solutions which demonstrate the versatility of the analytical model are described. (author)

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

Analysis of Environmental Data During TRIGA 2000 Reactor Operation in PTNBR-BATAN Bandung

International Nuclear Information System (INIS)

Zulfakhri

2009-01-01

The radioactivity data of grass and soil obtained from environmental monitoring during 2000-2008 have been quantitatively and qualitatively analyzed. The data were analyzed using statistical method of the population homogeneity varians test and the average equality test, whereas qualitative analysis have been carried out by gamma spectrometry. From the data analysis, using homogeneity varians test (X 2 test) and average equality test (F test) it was obtained that X 2 calculation , F calculation > X 2 table , F table , so it can be assumed that operation of TRIGA 2000 reactor has no effect to the radioactivity of the environment. The qualitative analysis of grass and soil samples shows no visible peak of gamma ray from fission or activation products from the reactor, but the natural radionuclide such as 214 Bi, 228 Ac, 212 Bi, and 40 K were detected. It can be concluded that reactor operation have no effect to the environmental radioactivity. (author)

Determination of a test section parameters for Iris nuclear reactor pressurizer

International Nuclear Information System (INIS)

Silva, Mario A.B. da; Lira, Carlos A.B. de O.

2009-01-01

An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered. (author)

The development of a small inherently safe homogeneous reactor for the production of medical isotopes

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)

Development of physical conceptions of fast reactors

International Nuclear Information System (INIS)

Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

2013-01-01

• Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

On the Evaluation of Pebble Bead Reactor Critical Experiments Using the Pebbed Code

International Nuclear Information System (INIS)

Gougar, Hans D.; Sen, R. Sonat

2014-01-01

Critical experiments pose a particular but necessary challenge to validating pebble bed reactor design codes. Fuel and core heterogeneities, impurities in graphite, variable packing of pebbles, and moderately strong neutronic coupling are among the factors that inject uncertainty into the results obtained with lower fidelity core physics models. Some of these are addressed in this study. The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling

Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

2012-07-15

Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

Darcy-Forchheimer flow with Cattaneo-Christov heat flux and homogeneous-heterogeneous reactions.

Science.gov (United States)

Hayat, Tasawar; Haider, Farwa; Muhammad, Taseer; Alsaedi, Ahmed

2017-01-01

Here Darcy-Forchheimer flow of viscoelastic fluids has been analyzed in the presence of Cattaneo-Christov heat flux and homogeneous-heterogeneous reactions. Results for two viscoelastic fluids are obtained and compared. A linear stretching surface has been used to generate the flow. Flow in porous media is characterized by considering the Darcy-Forchheimer model. Modified version of Fourier's law through Cattaneo-Christov heat flux is employed. Equal diffusion coefficients are employed for both reactants and auto catalyst. Optimal homotopy scheme is employed for solutions development of nonlinear problems. Solutions expressions of velocity, temperature and concentration fields are provided. Skin friction coefficient and heat transfer rate are computed and analyzed. Here the temperature and thermal boundary layer thickness are lower for Cattaneo-Christov heat flux model in comparison to classical Fourier's law of heat conduction. Moreover, the homogeneous and heterogeneous reactions parameters have opposite behaviors for concentration field.

Reactor cost driving items

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

RHEIN, Modular System for Reactor Design Calculation

International Nuclear Information System (INIS)

Reiche, Christian; Barz, Hansulrich; Kunzmann, Bernd; Seifert, Eberhard; Wand, Hartmut

1990-01-01

1 - Description of program or function: RHEIN is a modular reactor code system for neutron physics calculations. It consists of a small number of system codes for execution control, data management, and handling support, as well as of the physical calculation routines. The execution is controlled by input data containing mathematical and physical parameters and simple commands for routine calls and data manipulations. The calculation routines are in tune with one another and the system takes care of the data transfer between them. Cross-section libraries with self shielding parameters are added to the system. 2 - Method of solution: The calculation routines can be used for solving the following physics problems: - Calculation of cross-section sets for infinite mediums, taking into account chord length. - Zero-dimensional spectrum calculation in diffusion, P1, or B1 approximation. - One-dimensional calculation in diffusion, P1, or collision probability approximation. - Two-dimensional diffusion calculation. - Cell calculation by THERMOS. - Zone-wise homogenized group collapsing within zero, one, or two-dimensional models. - Normalization, summarizing, etc. - Output of cross-section sets to off systems Sn and Monte-Carlo calculations

Experimental investigation of the MSFR molten salt reactor concept

Energy Technology Data Exchange (ETDEWEB)

Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2014-11-15

In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

SEPARATION OF HYDROGEN AND CARBON DIOXIDE USING A NOVEL MEMBRANE REACTOR IN ADVANCED FOSSIL ENERGY CONVERSION PROCESS

Energy Technology Data Exchange (ETDEWEB)

Shamsuddin Ilias

2005-02-03

Inorganic membrane reactors offer the possibility of combining reaction and separation in a single operation at high temperatures to overcome the equilibrium limitations experienced in conventional reactor configurations. Such attractive features can be advantageously utilized in a number of potential commercial opportunities, which include dehydrogenation, hydrogenation, oxidative dehydrogenation, oxidation and catalytic decomposition reactions. However, to be cost effective, significant technological advances and improvements will be required to solve several key issues which include: (a) permselective thin solid film, (b) thermal, chemical and mechanical stability of the film at high temperatures, and (c) reactor engineering and module development in relation to the development of effective seals at high temperature and high pressure. In this project, we are working on the development and application of palladium and palladium-silver alloy thin-film composite membranes in membrane reactor-separator configuration for simultaneous production and separation of hydrogen and carbon dioxide at high temperature. From our research on Pd-composite membrane, we have demonstrated that the new membrane has significantly higher hydrogen flux with very high perm-selectivity than any of the membranes commercially available. The steam reforming of methane by equilibrium shift in Pd-composite membrane reactor is being studied to demonstrate the potential application of this new development. A two-dimensional, pseudo-homogeneous membrane-reactor model was developed to investigate the steam-methane reforming (SMR) reactions in a Pd-based membrane reactor. Radial diffusion was taken into consideration to account for the concentration gradient in the radial direction due to hydrogen permeation through the membrane. With appropriate reaction rate expressions, a set of partial differential equations was derived using the continuity equation for the reaction system. The equations were

There is no non-zero stable fixed point for dense networks in the homogeneous Kuramoto model

International Nuclear Information System (INIS)

Taylor, Richard

2012-01-01

This paper is concerned with the existence of multiple stable fixed point solutions of the homogeneous Kuramoto model. We develop a necessary condition for the existence of stable fixed points for the general network Kuramoto model. This condition is applied to show that for sufficiently dense n-node networks, with node degrees at least 0.9395(n−1), the homogeneous (equal frequencies) model has only one stable fixed point solution over the full space of phase angles in the range −π to π. This is the zero fixed point solution defined by all phase angle differences being zero. This result, together with existing research, proves a conjecture of Verwoerd and Mason (2007 Proc. of the American Control Conf. pp 4613–8) that for the complete network and the homogeneous model, the zero fixed point has a basin of attraction consisting of the entire space minus a set of measure zero. The necessary conditions are also tested to see how close to sufficiency they might be by applying them to a class of regular degree networks studied by Wiley et al (2006 Chaos 16 015103). (paper)

Neutronic study of a nuclear reactor of fused salts; Estudio neutronico de un reactor nuclear de sales fundidas

Energy Technology Data Exchange (ETDEWEB)

Garcia B, F. B.; Francois L, J. L., E-mail: faviolabelen@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2012-10-15

The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

Paracantor: A two group, two region reactor code

Energy Technology Data Exchange (ETDEWEB)

Stone, Stuart

1956-07-01

Paracantor I a two energy group, two region, time independent reactor code, which obtains a closed solution for a critical reactor assembly. The code deals with cylindrical reactors of finite length and with a radial reflector of finite thickness. It is programmed for the 1.B.M: Magnetic Drum Data-Processing Machine, Type 650. The limited memory space available does not permit a flux solution to be included in the basic Paracantor code. A supplementary code, Paracantor 11, has been programmed which computes fluxes, .including adjoint fluxes, from the .output of Paracamtor I.

Micro segregation and homogenization treatments of uranium-niobium alloys (U-Nb)

International Nuclear Information System (INIS)

Leal, Jose Fernando

1988-01-01

In the following sections micro segregation results in 0-3,6 wt% Nb and U-6,1 wt% Nb alloys casted in no consumable electrode arc furnace are presented. The micro segregation is studied qualitatively by optical microscopy and quantitatively by electron microprobe. The degree of homogenization has been measured after 800 and 850 deg C heat treatments in tubular resistive furnace. The microstructures after heat treatments are quantitatively analysed to check effects on the casting structures, mainly the variations in solute along the dendrite arm spacing. Some solidification phenomena are then discussed on reference to theoretical models of dendritic solidification , including microstructure and micro segregation. The experimental results are compared to theoretical on basis of initial and residual micro segregation after homogenization treatments. The times required for homogenization of the alloys are also discussed in function of the micro segregation from casting structures and the temperatures of the treatments. (author)

Reactor shutdown back-up system

International Nuclear Information System (INIS)

Hirao, Seizo; Sakashita, Motoaki.

1982-01-01

Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

Analysis of dynamic stability and safety of reactor system by reactor simulator; Analiza dinamicke stabilnosti i sigurnosti reaktorskog sistema pomocu reaktorskog simulatora

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1963-11-15

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.

Homogeneous and Heterogeneous (Fex, Cr1-x)(OH)3 Precipitation: Implications for Cr Sequestration.

Science.gov (United States)

Dai, Chong; Zuo, Xiaobing; Cao, Bo; Hu, Yandi

2016-02-16

The formation of (Fe, Cr)(OH)3 nanoparticles determines the fate of aqueous Cr in many aquatic environments. Using small-angle X-ray scattering, precipitation rates of (Fe, Cr)(OH)3 nanoparticles in solution and on quartz were quantified from 0.1 mM Fe(III) solutions containing 0-0.25 mM Cr(III) at pH = 3.7 ± 0.2. Concentration ratio of aqueous Cr(III)/Fe(III) controlled the chemical composition (x) of (Fex, Cr1-x)(OH)3 precipitates, solutions' supersaturation with respect to precipitates, and the surface charge of quartz. Therefore, the aqueous Cr(III)/Fe(III) ratio affected homogeneous (in solution) and heterogeneous (on quartz) precipitation rates of (Fex, Cr1-x)(OH)3 through different mechanisms. The sequestration mechanisms of Cr(III) in precipitates were also investigated. In solutions with high aqueous Cr(III)/Fe(III) ratios, surface enrichment of Cr(III) on the precipitates occurred, resulting in slower particle growth in solutions. From solutions with 0-0.1 mM Cr(III), the particles on quartz grew from 2 to 4 nm within 1 h. Interestingly, from solution with 0.25 mM Cr(III), particles of two distinct sizes (2 and 6 nm) formed on quartz, and their sizes remained unchanged throughout the reaction. Our study provided new insights on homogeneous and heterogeneous precipitation of (Fex, Cr1-x)(OH)3 nanoparticles, which can help determine the fate of Cr in aquatic environments.

The use of an equivalent homogeneous half-space in soil-structure interaction analysis

International Nuclear Information System (INIS)

Holzloehner, U.

1979-01-01

In analyses of seismic soil-structure interaction, the soil often is assumed as an elastic body. The solution procedure is lengthy if the heterogeneity of the soil is considered strictly. If the soil is taken as a homogeneous elastic half-space, existing solutions can be used. There are solutions for some simple layered systems, too. However, it is often not easy to correlate the variation of the soil properties with depth as found by measurements to those of ideal systems. The purpose of the paper is to show how to make use of the existing solutions. (orig.)

Spiked environmental matrix for use as a reference material for gamma-ray spectrometry: Production and homogeneity test

International Nuclear Information System (INIS)

Sobiech-Matura, K.; Máté, B.; Altzitzoglou, T.

2016-01-01

The application of a spiking method for reference material production and its utilisation for a food matrix is presented. The raw rice powder was tested by means of γ-ray spectrometry and spiked with a "1"3"7Cs solution. The spiked material was mixed and tested for homogeneity. The future use of the rice powder reference material after the entire characterisation cycle will be for γ-ray spectrometry method validation. - Highlights: • Spiking blank substance with a traceable radioactive solution • Spiked reference material for γ-ray emitting radionuclides in food matrix • Results of the homogeneity tests are presented

A bibliographic review of mathematical models of packed-bed biological reactors (PBR

Directory of Open Access Journals (Sweden)

Deisy Corredor

2005-09-01

Full Text Available Several authors have sublected packed-bed biological reactors to mathematical and theoretical analysis. They have taken reaction kinetics and single-dimensional, homogeneous, pseudo-homogeneous and heterogeneous models into account. Numerical methods have provided the set of equations so developed. The effect of physically important process variables in terms of design and operation have been investigated (i.e. residence time, operating- flow, substrate conversion, bio-film area and film thickness.

Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor

International Nuclear Information System (INIS)

Czernecki, Sebastien

1998-01-01

The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr

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

UWMAK-II: a conceptual tokamak reactor design

International Nuclear Information System (INIS)

1975-10-01

This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high

Multigrid Finite Element Method in Calculation of 3D Homogeneous and Composite Solids

Directory of Open Access Journals (Sweden)

A.D. Matveev

2016-12-01

Full Text Available In the present paper, a method of multigrid finite elements to calculate elastic three-dimensional homogeneous and composite solids under static loading has been suggested. The method has been developed based on the finite element method algorithms using homogeneous and composite three-dimensional multigrid finite elements (MFE. The procedures for construction of MFE of both rectangular parallelepiped and complex shapes have been shown. The advantages of MFE are that they take into account, following the rules of the microapproach, heterogeneous and microhomogeneous structures of the bodies, describe the three-dimensional stress-strain state (without any simplifying hypotheses in homogeneous and composite solids, as well as generate small dimensional discrete models and numerical solutions with a high accuracy.

Soybean epoxide production with in situ peracetic acid using homogeneous catalysis

Directory of Open Access Journals (Sweden)

Luis Alejandro Boyacá Mendivelso

2010-01-01

Full Text Available Using vegetable oils has become an excellent option for petrochemical product substitution. The epoxides obtained from such oils have wide applications as plastifiers and PVC stabilisers and as raw material in polyol synthesis for the polyurethane industry. This paper presents soybean oil epoxidation using a homogeneous catalyst in a well-mixed, stirred reactor being operated in iso- thermal conditions. The best result achieved was a 6.4% oxyrane oxygen content using hydrogen peroxide (25% molar excess, a- cetic acid (5% p/p and sulphuric acid (2% p/p concentrations at 80°C.

Fabrication of homogeneous titania/MWNT composite materials

International Nuclear Information System (INIS)

Korbely, Barbara; Nemeth, Zoltan; Reti, Balazs; Seo, Jin Won; Magrez, Arnaud; Forro, Laszlo; Hernadi, Klara

2011-01-01

Highlights: → Homogenous titania coverage on MWNT surface in a controllable way. → Various titanium alkoxy precursors are suitable for layer formation. → Acetone and ethanol are the best to promote interaction between MWNT and titania. -- Abstract: MWNT/titania nanocomposites were prepared by an impregnation method and subsequent heat treatment at 400 o C. Precursor compounds such as titanium (IV) propoxide and titanium (IV) ethoxide were used to cover the surface of CNTs under solution conditions. Electron microscopy and X-ray diffraction techniques were carried out to characterize the as-prepared titania layers.

Homogenization of Mammalian Cells.

Science.gov (United States)

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.

Reduction of Tc(VII) and Np(V) in solution by ferrous iron. A laboratory study of homogeneous and heterogeneous redox processes

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

Reduction of Tc(VII) and Np(V) in solution by ferrous iron. A laboratory study of homogeneous and heterogeneous redox processes

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.

Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

International Nuclear Information System (INIS)

Ferrufino, Felipe Bonito Jaldin

2011-01-01

IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

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

Preparation and properties of homogeneous-reinforced polyvinylidene fluoride hollow fiber membrane

Energy Technology Data Exchange (ETDEWEB)

Zhang Xuliang [State Key Laboratory of Hollow Fiber Membrane Materials and Processes, Tianjin Polytechnic University, Tianjin 300387 (China); Xiao Changfa, E-mail: xiaotjpu@163.com [State Key Laboratory of Hollow Fiber Membrane Materials and Processes, Tianjin Polytechnic University, Tianjin 300387 (China); Hu Xiaoyu; Bai Qianqian [State Key Laboratory of Hollow Fiber Membrane Materials and Processes, Tianjin Polytechnic University, Tianjin 300387 (China)

2013-01-01

Highlights: Black-Right-Pointing-Pointer The homogeneous-reinforced method has been adopted firstly in preparing of PVDF membranes. Black-Right-Pointing-Pointer The HR membranes have a favorable interfacial bonding between the coating layer and the matrix membrane. Black-Right-Pointing-Pointer The better performance of the HR membranes in protein solution can indirectly improve the service life of membranes. - Abstract: Homogeneous-reinforced (HR) polyvinylidene fluoride (PVDF) hollow fiber membranes include PVDF polymer solutions (coating layer) and the matrix membrane prepared through the dry-wet spinning process. The performance of HR membranes varies with the polymer concentration in the polymer solutions and is characterized in terms of pure water flux, rejection of protein, porosity, infiltration property, a mechanical strength test, and morphology observations by a field emission scanning electron microscope (FESEM). The results of this study indicate that the tensile strength of the HR PVDF membranes decreases slights compared with that of the matrix membrane, but the elongation at break increases much more and the hollow fiber membranes are endowed with better flexibility performance. The HR PVDF hollow fiber membranes have a favorable interfacial bonding between the coating layer and the matrix membrane, as shown by FESEM. The infiltration property is characterized by the contact angle experiments. Pure water flux decreases while the rejection ratio with an increase in polymer concentration increasing. The protein solution flux of the HR PVDF membranes is higher than that of the matrix membrane after 100 min of infiltration.

Preparation and properties of homogeneous-reinforced polyvinylidene fluoride hollow fiber membrane

International Nuclear Information System (INIS)

Zhang Xuliang; Xiao Changfa; Hu Xiaoyu; Bai Qianqian

2013-01-01

Highlights: ► The homogeneous-reinforced method has been adopted firstly in preparing of PVDF membranes. ► The HR membranes have a favorable interfacial bonding between the coating layer and the matrix membrane. ► The better performance of the HR membranes in protein solution can indirectly improve the service life of membranes. - Abstract: Homogeneous-reinforced (HR) polyvinylidene fluoride (PVDF) hollow fiber membranes include PVDF polymer solutions (coating layer) and the matrix membrane prepared through the dry-wet spinning process. The performance of HR membranes varies with the polymer concentration in the polymer solutions and is characterized in terms of pure water flux, rejection of protein, porosity, infiltration property, a mechanical strength test, and morphology observations by a field emission scanning electron microscope (FESEM). The results of this study indicate that the tensile strength of the HR PVDF membranes decreases slights compared with that of the matrix membrane, but the elongation at break increases much more and the hollow fiber membranes are endowed with better flexibility performance. The HR PVDF hollow fiber membranes have a favorable interfacial bonding between the coating layer and the matrix membrane, as shown by FESEM. The infiltration property is characterized by the contact angle experiments. Pure water flux decreases while the rejection ratio with an increase in polymer concentration increasing. The protein solution flux of the HR PVDF membranes is higher than that of the matrix membrane after 100 min of infiltration.

Final report on in-reactor uniaxial tensile deformation of pure iron and Fe-Cr alloy

DEFF Research Database (Denmark)

Singh, Bachu Narain; Huang, X.; Tähtinen, S.

, the in-reactor deformation leads to accumulation of dislocations in a homogeneous fashion and only to a modest density. No dislocation cells are formed during the in-reactor or post-irradiation deformation of Fe-Cr and pure iron. Furthermore, in both cases, the slip systems even in the planes with Schmid...... factor value of almost zero get activated during the in-reactor as well as post-irradiation deformation. The main implications of these results are briefly discussed....

Functionality and homogeneity.

NARCIS (Netherlands)

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,