FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback
International Nuclear Information System (INIS)
Shober, R.A.; Daly, T.A.; Ferguson, D.R.
1978-10-01
FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600
Seismically constrained two-dimensional crustal thermal structure of ...
Indian Academy of Sciences (India)
The temperature field within the crust is closely related to tectonic history as well as many other geological processes inside the earth. Therefore, knowledge of the crustal thermal structure of a region is of great importance for its tectonophysical studies. This work deals with the two-dimensional thermal modelling to ...
Two-dimensional disruption thermal analysis code DREAM
International Nuclear Information System (INIS)
Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.
1988-08-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)
Thermal expansion of two-dimensional itinerant nearly ferromagnetic metal
International Nuclear Information System (INIS)
Konno, R; Hatayama, N; Takahashi, Y; Nakano, H
2009-01-01
Thermal expansion of two-dimensional itinerant nearly ferromagnetic metal is investigated according to the recent theoretical development of magneto-volume effect for the three-dimensional weak ferromagnets. We particularly focus on the T 2 -linear thermal expansion of magnetic origin at low temperatures, so far disregarded by conventional theories. As the effect of thermal spin fluctuations we have found that the T-linear thermal expansion coefficient shows strong enhancement by assuming the double Lorentzian form of the non-interacting dynamical susceptibility justified in the small wave-number and low frequency region. It grows faster in proportional to y -1/2 as we approach the magnetic instability point than two-dimensional nearly antiferromagnetic metals with ln(1/y s ) dependence, where y and y s are the inverses of the reduced uniform and staggered magnetic susceptibilities, respectively. Our result is consistent with the Grueneisen's relation between the thermal expansion coefficient and the specific heat at low temperatures. In 2-dimensional electron gas we find that the thermal expansion coefficient is divergent with a finite y when the higher order term of non-interacting dynamical susceptibility is taken into account.
Two-dimensional fruit ripeness estimation using thermal imaging
Sumriddetchkajorn, Sarun; Intaravanne, Yuttana
2013-06-01
Some green fruits do not change their color from green to yellow when being ripe. As a result, ripeness estimation via color and fluorescent analytical approaches cannot be applied. In this article, we propose and show for the first time how a thermal imaging camera can be used to two-dimensionally classify fruits into different ripeness levels. Our key idea relies on the fact that the mature fruits have higher heat capacity than the immature ones and therefore the change in surface temperature overtime is slower. Our experimental proof of concept using a thermal imaging camera shows a promising result in non-destructively identifying three different ripeness levels of mangoes Mangifera indica L.
Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis
International Nuclear Information System (INIS)
Downar, T.J.
1987-01-01
A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback
Two-dimensional thermal analysis of liquid hydrogen tank insulation
Energy Technology Data Exchange (ETDEWEB)
Babac, Gulru; Sisman, Altug [Istanbul Technical University, Energy Institute, Ayazaga campus, 34469 Maslak, Istanbul (Turkey); Cimen, Tolga [Jaguar and Landrover, Banbury Road, Gaydon, Warwick CV35 0RR (United Kingdom)
2009-08-15
Liquid hydrogen (LH{sub 2}) storage has the advantage of high volumetric energy density, while boil-off losses constitute a major disadvantage. To minimize the losses, complicated insulation techniques are necessary. In general, Multi Layer Insulation (MLI) and a Vapor-Cooled Shield (VCS) are used together in LH{sub 2} tanks. In the design of an LH{sub 2} tank with VCS, the main goal is to find the optimum location for the VCS in order to minimize heat leakage. In this study, a 2D thermal model is developed by considering the temperature dependencies of the thermal conductivity and heat capacity of hydrogen gas. The developed model is used to analyze the effects of model considerations on heat leakage predictions. Furthermore, heat leakage in insulation of LH{sub 2} tanks with single and double VCS is analyzed for an automobile application, and the optimum locations of the VCS for minimization of heat leakage are determined for both cases. (author)
Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling
Domalapally, Phani; Di Caro, Marco
2018-05-01
Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.
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.
A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor
International Nuclear Information System (INIS)
Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.
1988-01-01
A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab
Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor
International Nuclear Information System (INIS)
Karpov, V.A.; Protsenko, A.N.
1975-01-01
Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)
International Nuclear Information System (INIS)
Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.
2000-01-01
The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively
International Nuclear Information System (INIS)
Abreu, M.P. de.
1988-01-01
An alternative pseudo-harmonics method for two-dimensional reactor calculations is presented together with some one-energy group results, namely, eigenvalue and flux reconstruction. A brief description of the Standard and Modified versions of the method is presented for critical purposes, i.e., it was intended to discuss the previously developed versions and in some sense to improve the solution of the K-th eigenvalue and flux terms of the corresponding expansions. Intense and localized perturbations, where a significant imbalance between neutron production and destruction rates exists, were simulated. Since convergence in flux and eigenvalue were achieved for all test-cases, there is a tendency to consider the alternative method to be very promising for two-dimensional calculations. (author)
DEFF Research Database (Denmark)
Swierczynski, Maciej Jozef; Stroe, Daniel Loan; Knap, Vaclav
2016-01-01
Thermal modeling of lithium-ion batteries is gaining its importance together with increasing power density and compact design of the modern battery systems in order to assure battery safety and long lifetime. Thermal models of lithium-ion batteries are usually either expensive to develop...... and accurate or equivalent thermal circuit based with moderate accuracy and without spatial temperature distribution. This work presents initial results that can be used as a fundament for the cost-efficient development of the two-dimensional thermal model of lithium-ion battery based on multipoint...
Application of synthesis methods to two-dimensional fast reactor transient study
International Nuclear Information System (INIS)
Izutsu, Sadayuki; Hirakawa, Naohiro
1978-01-01
Space time synthesis and time synthesis codes were developed and applied to the space-dependent kinetics benchmark problem of a two-dimensional fast reactor model, and it was found both methods are accurate and economical for the fast reactor kinetics study. Comparison between the space time synthesis and the time synthesis was made. Also, in space time synthesis, the influence of the number of trial functions on the error and on the computing time and the effect of degeneration of expansion coefficients are investigated. The matrix factorization method is applied to the inversion of the matrix equation derived from the synthesis equation, and it is indicated that by the use of this scheme space-dependent kinetics problem of a fast reactor can be solved efficiently by space time synthesis. (auth.)
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.
Test of quantum thermalization in the two-dimensional transverse-field Ising model.
Blaß, Benjamin; Rieger, Heiko
2016-12-01
We study the quantum relaxation of the two-dimensional transverse-field Ising model after global quenches with a real-time variational Monte Carlo method and address the question whether this non-integrable, two-dimensional system thermalizes or not. We consider both interaction quenches in the paramagnetic phase and field quenches in the ferromagnetic phase and compare the time-averaged probability distributions of non-conserved quantities like magnetization and correlation functions to the thermal distributions according to the canonical Gibbs ensemble obtained with quantum Monte Carlo simulations at temperatures defined by the excess energy in the system. We find that the occurrence of thermalization crucially depends on the quench parameters: While after the interaction quenches in the paramagnetic phase thermalization can be observed, our results for the field quenches in the ferromagnetic phase show clear deviations from the thermal system. These deviations increase with the quench strength and become especially clear comparing the shape of the thermal and the time-averaged distributions, the latter ones indicating that the system does not completely lose the memory of its initial state even for strong quenches. We discuss our results with respect to a recently formulated theorem on generalized thermalization in quantum systems.
Richter, Johannes M; Branchi, Federico; Valduga de Almeida Camargo, Franco; Zhao, Baodan; Friend, Richard H; Cerullo, Giulio; Deschler, Felix
2017-08-29
In band-like semiconductors, charge carriers form a thermal energy distribution rapidly after optical excitation. In hybrid perovskites, the cooling of such thermal carrier distributions occurs on timescales of about 300 fs via carrier-phonon scattering. However, the initial build-up of the thermal distribution proved difficult to resolve with pump-probe techniques due to the requirement of high resolution, both in time and pump energy. Here, we use two-dimensional electronic spectroscopy with sub-10 fs resolution to directly observe the carrier interactions that lead to a thermal carrier distribution. We find that thermalization occurs dominantly via carrier-carrier scattering under the investigated fluences and report the dependence of carrier scattering rates on excess energy and carrier density. We extract characteristic carrier thermalization times from below 10 to 85 fs. These values allow for mobilities of 500 cm 2 V -1 s -1 at carrier densities lower than 2 × 10 19 cm -3 and limit the time for carrier extraction in hot carrier solar cells.Carrier-carrier scattering rates determine the fundamental limits of carrier transport and electronic coherence. Using two-dimensional electronic spectroscopy with sub-10 fs resolution, Richter and Branchi et al. extract carrier thermalization times of 10 to 85 fs in hybrid perovskites.
Test of quantum thermalization in the two-dimensional transverse-field Ising model
Blaß, Benjamin; Rieger, Heiko
2016-01-01
We study the quantum relaxation of the two-dimensional transverse-field Ising model after global quenches with a real-time variational Monte Carlo method and address the question whether this non-integrable, two-dimensional system thermalizes or not. We consider both interaction quenches in the paramagnetic phase and field quenches in the ferromagnetic phase and compare the time-averaged probability distributions of non-conserved quantities like magnetization and correlation functions to the thermal distributions according to the canonical Gibbs ensemble obtained with quantum Monte Carlo simulations at temperatures defined by the excess energy in the system. We find that the occurrence of thermalization crucially depends on the quench parameters: While after the interaction quenches in the paramagnetic phase thermalization can be observed, our results for the field quenches in the ferromagnetic phase show clear deviations from the thermal system. These deviations increase with the quench strength and become especially clear comparing the shape of the thermal and the time-averaged distributions, the latter ones indicating that the system does not completely lose the memory of its initial state even for strong quenches. We discuss our results with respect to a recently formulated theorem on generalized thermalization in quantum systems. PMID:27905523
Sirmas, Nick; Radulescu, Matei I.
2016-01-01
The problem of thermal ignition in a homogeneous gas is revisited from a molecular dynamics perspective. A two-dimensional model is adopted, which assumes reactive disks of type A and B in a fixed area that react to form type C products if an activation threshold for impact is surpassed. Such a reaction liberates kinetic energy to the product particles, representative of the heat release. The results for the ignition delay are compared with those obtained from the continuum description assumi...
Two-dimensional nucleonics calculations for a ''FIRST STEP'' conceptual ICF reactor
International Nuclear Information System (INIS)
Davidson, J.W.; Battat, M.E.; Saylor, W.W.; Pendergrass, J.H.; Dudziak, D.J.
1985-01-01
A detailed two-dimensional nucleonic analysis has been performed for the FIRST STEP conceptual ICF reactor blanket design. The reactor concept incorporated in this design is a modified wetted-wall cavity with target illumination geometry left as a design variable. The 2-m radius spherical cavity is surrounded by a blanket containing lithium and 238 U as fertile species and also as energy multipliers. The blanket is configured as 0.6-m-thick cylindrical annuli containing modified LMFBR-type fuel elements with 0.5-m-thick fuel-bearing axial end plugs. Liquid lithium surrounds the inner blanket regions and serves as the coolant for both the blanket and the first wall. The two-dimensional analysis of the blanket performance was made using the 2-D discrete-ordinates code TRISM, and benchmarked with the 3-D Monte Carlo code MCNP. Integral responses including the tritium breeding ratio (TBR), plutonium breeding ratio (PUBR), and blanket energy multiplication were calculated for axial and radial blanket regions. Spatial distributions were calculated for steady-state rates of fission, neutron heating, prompt gamma-ray heating, and fuel breeding
Longitudinal On-Column Thermal Modulation for Comprehensive Two-Dimensional Liquid Chromatography.
Creese, Mari E; Creese, Mathew J; Foley, Joe P; Cortes, Hernan J; Hilder, Emily F; Shellie, Robert A; Breadmore, Michael C
2017-01-17
Longitudinal on-column thermal modulation for comprehensive two-dimensional liquid chromatography is introduced. Modulation optimization involved a systematic investigation of heat transfer, analyte retention, and migration velocity at a range of temperatures. Longitudinal on-column thermal modulation was realized using a set of alkylphenones and compared to a conventional valve-modulator employing sample loops. The thermal modulator showed a reduced modulation-induced pressure impact than valve modulation, resulting in reduced baseline perturbation by a factor of 6; yielding a 6-14-fold improvement in signal-to-noise. A red wine sample was analyzed to demonstrate the potential of the longitudinal on-column thermal modulator for separation of a complex sample. Discrete peaks in the second dimension using the thermal modulator were 30-55% narrower than with the valve modulator. The results shown herein demonstrate the benefits of an active focusing modulator, such as reduced detection limits and increased total peak capacity.
Numerical simulations of thermal conductivity in dissipative two-dimensional Yukawa systems.
Khrustalyov, Yu V; Vaulina, O S
2012-04-01
Numerical data on the heat transfer constants in two-dimensional Yukawa systems were obtained. Numerical study of the thermal conductivity and diffusivity was carried out for the equilibrium systems with parameters close to conditions of laboratory experiments with dusty plasma. For calculations of heat transfer constants the Green-Kubo formulas were used. The influence of dissipation (friction) on the heat transfer processes in nonideal systems was investigated. The approximation of the coefficient of thermal conductivity is proposed. Comparison of the obtained results to the existing experimental and numerical data is discussed.
A two dimensional approach for temperature distribution in reactor lower head during severe accident
International Nuclear Information System (INIS)
Cao, Zhen; Liu, Xiaojing; Cheng, Xu
2015-01-01
Highlights: • Two dimensional module is developed to analyze integrity of lower head. • Verification step has been done to evaluate feasibility of new module. • The new module is applied to simulate large-scale advanced PWR. • Importance of 2-D approach is clearly quantified. • Major parameters affecting vessel temperature distribution are identified. - Abstract: In order to evaluate the safety margin during a postulated severe accident, a module named ASAP-2D (Accident Simulation on Pressure vessel-2 Dimensional), which can be implemented into the severe accident simulation codes (such as ATHLET-CD), is developed in Shanghai Jiao Tong University. Based on two-dimensional spherical coordinates, heat conduction equation for transient state is solved implicitly. Together with solid vessel thickness, heat flux distribution and heat transfer coefficient at outer vessel surface are obtained. Heat transfer regime when critical heat flux has been exceeded (POST-CHF regime) could be simulated in the code, and the transition behavior of boiling crisis (from spatial and temporal points of view) can be predicted. The module is verified against a one-dimensional analytical solution with uniform heat flux distribution, and afterwards this module is applied to the benchmark illustrated in NUREG/CR-6849. Benchmark calculation indicates that maximum heat flux at outer surface of RPV could be around 20% lower than that of at inner surface due to two-dimensional heat conduction. Then a preliminary analysis is performed on the integrity of the reactor vessel for which the geometric parameters and boundary conditions are derived from a large scale advanced pressurized water reactor. Results indicate that heat flux remains lower than critical heat flux. Sensitivity analysis indicates that outer heat flux distribution is more sensitive to input heat flux distribution and the transition boiling correlation than mass flow rate in external reactor vessel cooling (ERVC) channel
Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis
International Nuclear Information System (INIS)
Arien, B.; Daniels, J.
1986-12-01
CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)
Thermal transport in a two-dimensional Z2 spin liquid
Metavitsiadis, Alexandros; Pidatella, Angelo; Brenig, Wolfram
2017-11-01
We study the dynamical thermal conductivity of the two-dimensional Kitaev spin model on the honeycomb lattice. We find a strongly temperature dependent low-frequency spectral intensity as a direct consequence of fractionalization of spins into mobile Majorana matter and a static Z2 gauge field. The latter acts as an emergent thermally activated disorder, leading to the appearance of a pseudogap which closes in the thermodynamic limit, indicating a dissipative heat conductor. Our analysis is based on complementary calculations of the current correlation function, comprising exact diagonalization by means of a complete summation over all gauge sectors, as well as a phenomenological mean-field treatment of thermal gauge fluctuations, valid at intermediate and high temperatures. The results will also be contrasted against the conductivity discarding gauge fluctuations.
Thermal structure of the ionosphere of Mars - simulations with one- and two-dimensional models
International Nuclear Information System (INIS)
Singhal, R.P.; Whitten, R.C.
1988-01-01
Heat flux saturation effects are included in the present one- and two-dimensional models of the Martian upper ionosphere's thermal structure. The inclusion of small upper boundary and volume heat sources is found to yield satisfactory simulations of the dayside ion temperature observation results obtained by Viking 1's retarding potential analyzers. It is noted that the plasma flow-transport of heat from the dayside to the nightside makes no contribution to the ion and electron temperatures that have been calculated for the nightside. 22 references
The band gap variation of a two dimensional binary locally resonant structure in thermal environment
Directory of Open Access Journals (Sweden)
Zhen Li
2017-01-01
Full Text Available In this study, the numerical investigation of thermal effect on band gap dynamical characteristic for a two-dimensional binary structure composed of aluminum plate periodically filled with nitrile rubber cylinder is presented. Initially, the band gap of the binary structure variation trend with increasing temperature is studied by taking the softening effect of thermal stress into account. A breakthrough is made which found the band gap being narrower and shifting to lower frequency in thermal environment. The complete band gap which in higher frequency is more sensitive to temperature that it disappears with temperature increasing. Then some new transformed models are created by changing the height of nitrile rubber cylinder from 1mm to 7mm. Simulations show that transformed model can produce a wider band gap (either flexure or complete band gap. A proper forbidden gap of elastic wave can be utilized in thermal environment although both flexure and complete band gaps become narrower with temperature. Besides that, there is a zero-frequency flat band appearing in the first flexure band, and it becomes broader with temperature increasing. The band gap width decreases trend in thermal environment, as well as the wider band gap induced by the transformed model with higher nitrile rubber cylinder is useful for the design and application of phononic crystal structures in thermal environment.
Thermal conductivity of a two-dimensional phosphorene sheet: a comparative study with graphene.
Hong, Yang; Zhang, Jingchao; Huang, Xiaopeng; Zeng, Xiao Cheng
2015-11-28
A recently discovered two-dimensional (2D) layered material phosphorene has attracted considerable interest as a promising p-type semiconducting material. In this work, thermal conductivity (κ) of monolayer phosphorene is calculated using large-scale classical non-equilibrium molecular dynamics (NEMD) simulations. The predicted thermal conductivities for infinite length armchair and zigzag phosphorene sheets are 63.6 and 110.7 W m(-1) K(-1) respectively. The strong anisotropic thermal transport is attributed to the distinct atomic structures at altered chiral directions and direction-dependent group velocities. Thermal conductivities of 2D graphene sheets with the same dimensions are also computed for comparison. The extrapolated κ of the 2D graphene sheet are 1008.5(+37.6)(-37.6) and 1086.9(+59.1)(-59.1) W m(-1) K(-1) in the armchair and zigzag directions, respectively, which are an order of magnitude higher than those of phosphorene. The overall and decomposed phonon density of states (PDOS) are calculated in both structures to elucidate their thermal conductivity differences. In comparison with graphene, the vibrational frequencies that can be excited in phosphorene are severely limited. The temperature effect on the thermal conductivity of phosphorene and graphene sheets is investigated, which reveals a monotonic decreasing trend for both structures.
Two-dimensional transient thermal analysis of a fuel rod by finite volume method
Energy Technology Data Exchange (ETDEWEB)
Costa, Rhayanne Yalle Negreiros; Silva, Mário Augusto Bezerra da; Lira, Carlos Alberto de Oliveira, E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear
2017-07-01
One of the greatest concerns when studying a nuclear reactor is the warranty of safe temperature limits all over the system at all time. The preservation of core structure along with the constraint of radioactive material into a controlled system are the main focus during the operation of a reactor. The purpose of this paper is to present the temperature distribution for a nominal channel of the AP1000 reactor developed by Westinghouse Co. during steady-state and transient operations. In the analysis, the system was subjected to normal operation conditions and then to blockages of the coolant flow. The time necessary to achieve a new safe stationary stage (when it was possible) was presented. The methodology applied in this analysis was based on a two-dimensional survey accomplished by the application of Finite Volume Method (FVM). A steady solution is obtained and compared with an analytical analysis that disregard axial heat transport to determine its relevance. The results show the importance of axial heat transport consideration in this type of study. A transient analysis shows the behavior of the system when submitted to coolant blockage at channel's entrance. Three blockages were simulated (10%, 20% and 30%) and the results show that, for a nominal channel, the system can still be considerate safe (there's no bubble formation until that point). (author)
Thermality and excited state Rényi entropy in two-dimensional CFT
Energy Technology Data Exchange (ETDEWEB)
Lin, Feng-Li [Department of Physics, National Taiwan Normal University,Taipei 11677, Taiwan (China); Wang, Huajia [Department of Physics, University of Illinois,Urbana-Champaign, IL 61801 (United States); Zhang, Jia-ju [Dipartimento di Fisica, Università degli Studi di Milano-Bicocca,Piazza della Scienza 3, I-20126 Milano (Italy); Theoretical Physics Division, Institute of High Energy Physics, Chinese Academy of Sciences,19B Yuquan Rd, Beijing 100049 (China); Theoretical Physics Center for Science Facilities, Chinese Academy of Sciences,19B Yuquan Rd, Beijing 100049 (China)
2016-11-21
We evaluate one-interval Rényi entropy and entanglement entropy for the excited states of two-dimensional conformal field theory (CFT) on a cylinder, and examine their differences from the ones for the thermal state. We assume the interval to be short so that we can use operator product expansion (OPE) of twist operators to calculate Rényi entropy in terms of sum of one-point functions of OPE blocks. We find that the entanglement entropy for highly excited state and thermal state behave the same way after appropriate identification of the conformal weight of the state with the temperature. However, there exists no such universal identification for the Rényi entropy in the short-interval expansion. Therefore, the highly excited state does not look thermal when comparing its Rényi entropy to the thermal state one. As the Rényi entropy captures the higher moments of the reduced density matrix but the entanglement entropy only the average, our results imply that the emergence of thermality depends on how refined we look into the entanglement structure of the underlying pure excited state.
COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests
International Nuclear Information System (INIS)
Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi
1987-01-01
The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)
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)
Atomic reactor thermal engineering
International Nuclear Information System (INIS)
Kim, Gwang Ryong
1983-02-01
This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.
Two-dimensional simulation of the thermal stress effect on static and dynamic VDMOS characteristics
International Nuclear Information System (INIS)
Alwan, M.; Beydoun, B.; Ketata, K.; Zoaeter, M.
2005-01-01
Using a two-dimensional simulator, the effect of the thermal stress on static and dynamic vertical double-diffusion metal oxide semiconductor (VDMOS) characteristics have been investigated. The use of the device under certain thermal stress conditions can produce modifications of its physical and electrical properties. Based on physics and 2D simulations, this paper proposes an analysis of this stress effect observed on the electrical characteristics of the device. Parameters responsible of these modifications are determined. Approximate expressions of the ionization coefficients and breakdown voltage in terms of temperature are proposed. Non-punch-through junction theory is used to express the breakdown voltage and the space charge extension with respect to the impurity concentration and the temperature. The capacitances of the device have been also studied. The effect of the stress on C-V characteristics is observed and analyzed. We notice that the drain-gate, drain-source and gate-source capacitances are shifted due to the degradation of device physical properties versus thermal stress
Yin, Zhifu; Sun, Lei; Zou, Helin; Cheng, E.
2015-05-01
A method for obtaining a low-cost and high-replication precision two-dimensional (2D) nanofluidic device with a polymethyl methacrylate (PMMA) sheet is proposed. To improve the replication precision of the 2D PMMA nanochannels during the hot embossing process, the deformation of the PMMA sheet was analyzed by a numerical simulation method. The constants of the generalized Maxwell model used in the numerical simulation were calculated by experimental compressive creep curves based on previously established fitting formula. With optimized process parameters, 176 nm-wide and 180 nm-deep nanochannels were successfully replicated into the PMMA sheet with a replication precision of 98.2%. To thermal bond the 2D PMMA nanochannels with high bonding strength and low dimensional loss, the parameters of the oxygen plasma treatment and thermal bonding process were optimized. In order to measure the dimensional loss of 2D nanochannels after thermal bonding, a dimension loss evaluating method based on the nanoindentation experiments was proposed. According to the dimension loss evaluating method, the total dimensional loss of 2D nanochannels was 6 nm and 21 nm in width and depth, respectively. The tensile bonding strength of the 2D PMMA nanofluidic device was 0.57 MPa. The fluorescence images demonstrate that there was no blocking or leakage over the entire microchannels and nanochannels.
Two-dimensional multigroup finite element calculation of fast reactor in diffusion approximation
International Nuclear Information System (INIS)
Schmid, J.
1986-06-01
When a linear element of triangular shape is used the actual finite element calculation is relatively simple. Extensive programs for mesh generation were written for easy inputting the configuration of reactors. A number of other programs were written for plotting neutron flux fields in individual groups, the power distribution, spatial plotting of fields, etc. The operation of selected programs, data preparation and operating instructions are described and examples given of data and results. All programs are written in GIER ALGOL. The used method and the developed programs have demonstrated that they are a useful instrument for the calculation of criticality and the distribution of neutron flux and power of both fast and thermal reactors. (J.B.)
Thermalization of a two-dimensional photonic gas in a `white wall' photon box
Klaers, Jan; Vewinger, Frank; Weitz, Martin
2010-07-01
Bose-Einstein condensation, the macroscopic accumulation of bosonic particles in the energetic ground state below a critical temperature, has been demonstrated in several physical systems. The perhaps best known example of a bosonic gas, blackbody radiation, however exhibits no Bose-Einstein condensation at low temperatures. Instead of collectively occupying the lowest energy mode, the photons disappear in the cavity walls when the temperature is lowered-corresponding to a vanishing chemical potential. Here we report on evidence for a thermalized two-dimensional photon gas with a freely adjustable chemical potential. Our experiment is based on a dye-filled optical microresonator, acting as a `white wall' box for photons. Thermalization is achieved in a photon-number-conserving way by photon scattering off the dye molecules, and the cavity mirrors provide both an effective photon mass and a confining potential-key prerequisites for the Bose-Einstein condensation of photons. As a striking example of the unusual system properties, we demonstrate a yet unobserved light concentration effect into the centre of the confining potential, an effect with prospects for increasing the efficiency of diffuse solar light collection.
Limitations to the use of two-dimensional thermal modeling of a nuclear waste repository
International Nuclear Information System (INIS)
Davis, B.W.
1979-01-01
Thermal modeling of a nuclear waste repository is basic to most waste management predictive models. It is important that the modeling techniques accurately determine the time-dependent temperature distribution of the waste emplacement media. Recent modeling studies show that the time-dependent temperature distribution can be accurately modeled in the far-field using a 2-dimensional (2-D) planar numerical model; however, the near-field cannot be modeled accurately enough by either 2-D axisymmetric or 2-D planar numerical models for repositories in salt. The accuracy limits of 2-D modeling were defined by comparing results from 3-dimensional (3-D) TRUMP modeling with results from both 2-D axisymmetric and 2-D planar. Both TRUMP and ADINAT were employed as modeling tools. Two-dimensional results from the finite element code, ADINAT were compared with 2-D results from the finite difference code, TRUMP; they showed almost perfect correspondence in the far-field. This result adds substantially to confidence in future use of ADINAT and its companion stress code ADINA for thermal stress analysis. ADINAT was found to be somewhat sensitive to time step and mesh aspect ratio. 13 figures, 4 tables
Two dimensional finite element thermal model of laser surface glazing for H13 tool steel
Kabir, I. R.; Yin, D.; Naher, S.
2016-10-01
A two dimensional (2D) transient thermal model with line-heat-source was developed by Finite Element Method (FEM) for laser surface glazing of H13 tool steel using commercial software-ANSYS 15. The geometry of the model was taken as a transverse circular cross-section of cylindrical specimen. Two different power levels (300W, 200W) were used with 0.2mm width of laser beam and 0.15ms exposure time. Temperature distribution, heating and cooling rates, and the dimensions of modified surface were analysed. The maximum temperatures achieved were 2532K (2259°C) and 1592K (1319°C) for laser power 300W and 200W respectively. The maximum cooling rates were 4.2×107 K/s for 300W and 2×107 K/s for 200W. Depths of modified zone increased with increasing laser power. From this analysis, it can be predicted that for 0.2mm beam width and 0.15ms time exposer melting temperature of H13 tool steel is achieved within 200-300W power range of laser beam in laser surface glazing.
Two-dimensional cross-section sensitivity and uncertainty analysis for fusion reactor blankets
International Nuclear Information System (INIS)
Embrechts, M.J.
1982-02-01
A two-dimensional sensitivity and uncertainty analysis for the heating of the TF coil for the FED (fusion engineering device) blanket was performed. The uncertainties calculated are of the same order of magnitude as those resulting from a one-dimensional analysis. The largest uncertainties were caused by the cross section uncertainties for chromium
International Nuclear Information System (INIS)
Wong, K.-L.; Hsien, T.-L.; Hsiao, M.-C.; Chen, W.-L.; Lin, K.-C.
2008-01-01
This investigation is to show that two-dimensional steady state heat transfer problems of composite walls should not be solved by the conventionally one-dimensional parallel thermal resistance circuits (PTRC) model because the interface temperatures are not unique. Thus PTRC model cannot be used like its conventional recognized analogy, parallel electrical resistance circuits (PERC) model which has the unique node electric voltage. Two typical composite wall examples, solved by CFD software, are used to demonstrate the incorrectness. The numerical results are compared with those obtained by PTRC model, and very large differences are observed between their results. This proves that the application of conventional heat transfer PTRC model to two-dimensional composite walls, introduced in most heat transfer text book, is totally incorrect. An alternative one-dimensional separately series thermal resistance circuit (SSTRC) model is proposed and applied to the two-dimensional composite walls with isothermal boundaries. Results with acceptable accuracy can be obtained by the new model
International Nuclear Information System (INIS)
Wong, K.-L.; Hsien, T.-L.; Chen, W.-L.; Yu, S.-J.
2008-01-01
This study is to prove that two-dimensional steady state heat transfer problems of composite circular pipes cannot be appropriately solved by the conventional one-dimensional parallel thermal resistance circuits (PTRC) model because its interface temperatures are not unique. Thus, the PTRC model is definitely different from its conventional recognized analogy, parallel electrical resistance circuits (PERC) model, which has unique node electric voltages. Two typical composite circular pipe examples are solved by CFD software, and the numerical results are compared with those obtained by the PTRC model. This shows that the PTRC model generates large error. Thus, this conventional model, introduced in most heat transfer text books, cannot be applied to two-dimensional composite circular pipes. On the contrary, an alternative one-dimensional separately series thermal resistance circuit (SSTRC) model is proposed and applied to a two-dimensional composite circular pipe with isothermal boundaries, and acceptable results are returned
International Nuclear Information System (INIS)
Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee
2009-01-01
The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect
Two-dimensional thermal modeling of power monolithic microwave integrated circuits (MMIC's)
Fan, Mark S.; Christou, Aris; Pecht, Michael G.
1992-01-01
Numerical simulations of the two-dimensional temperature distributions for a typical GaAs MMIC circuit are conducted, aiming at understanding the heat conduction process of the circuit chip and providing temperature information for device reliability analysis. The method used is to solve the two-dimensional heat conduction equation with a control-volume-based finite difference scheme. In particular, the effects of the power dissipation and the ambient temperature are examined, and the criterion for the worst operating environment is discussed in terms of the allowed highest device junction temperature.
International Nuclear Information System (INIS)
Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Torii, Yoshiya; Horiguchi, Yoji
2001-03-01
To evaluate nitrogen dose, boron dose and gamma-ray dose occurred by neutron capture reaction of the hydrogen at the medical irradiation, two-dimensional distribution of the thermal neutron flux is very important because these doses are proportional to the thermal neutron distribution. This report describes the measurement of the two-dimensional thermal neutron distribution in a head water phantom by neutron beams of the JRR-4 and evaluation of the dose distribution characteristic. Thermal neutron flux in the phantom was measured by gold wire placed in the spokewise of every 30 degrees in order to avoid the interaction. Distribution of the thermal neutron flux was also calculated using two-dimensional Lagrange's interpolation program (radius, angle direction) developed this time. As a result of the analysis, it was confirmed to become distorted distribution which has annular peak at outside of the void, though improved dose profile of the deep direction was confirmed in the case which the radiation field in the phantom contains void. (author)
Kim, Sung-Jin; Reidy, Shaelah M; Block, Bruce P; Wise, Kensall D; Zellers, Edward T; Kurabayashi, Katsuo
2010-07-07
In comprehensive two-dimensional gas chromatography (GC x GC), a modulator is placed at the juncture between two separation columns to focus and re-inject eluting mixture components, thereby enhancing the resolution and the selectivity of analytes. As part of an effort to develop a microGC x microGC prototype, in this report we present the design, fabrication, thermal operation, and initial testing of a two-stage microscale thermal modulator (microTM). The microTM contains two sequential serpentine Pyrex-on-Si microchannels (stages) that cryogenically trap analytes eluting from the first-dimension column and thermally inject them into the second-dimension column in a rapid, programmable manner. For each modulation cycle (typically 5 s for cooling with refrigeration work of 200 J and 100 ms for heating at 10 W), the microTM is kept approximately at -50 degrees C by a solid-state thermoelectric cooling unit placed within a few tens of micrometres of the device, and heated to 250 degrees C at 2800 degrees C s(-1) by integrated resistive microheaters and then cooled back to -50 degrees C at 250 degrees C s(-1). Thermal crosstalk between the two stages is less than 9%. A lumped heat transfer model is used to analyze the device design with respect to the rates of heating and cooling, power dissipation, and inter-stage thermal crosstalk as a function of Pyrex-membrane thickness, air-gap depth, and stage separation distance. Experimental results are in agreement with trends predicted by the model. Preliminary tests using a conventional capillary column interfaced to the microTM demonstrate the capability for enhanced sensitivity and resolution as well as the modulation of a mixture of alkanes.
A two-dimensional simulator of the neutronic behaviour of low power fast reactors
International Nuclear Information System (INIS)
Penha, M.A.V.R. da.
1984-01-01
A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt
Calculation of two-dimensional thermal transients by the finite element method
International Nuclear Information System (INIS)
Fontoura Rodrigues, J.L.A. da; Barcellos, C.S. de
1981-01-01
The linear heat conduction through anisotropic and/or heterogeneous matter, in either two-dimensional fields with any kind of geometry or three-dimensional fields with axial symmetry is analysed. It only accepts time-independent boundary conditions and it is possible to have internal heat generation. The solution is obtained by modal analysis employing the finite element method under Galerkin formulation. (Author) [pt
International Nuclear Information System (INIS)
Bezotosnyi, V V; Kumykov, Kh Kh
1998-01-01
A two-dimensional transient thermal model of an injection laser is developed. This model makes it possible to analyse the temperature profiles in pulsed and cw stripe lasers with an arbitrary width of the stripe contact, and also in linear laser-diode arrays. This can be done for any durations and repetition rates of the pump pulses. The model can also be applied to two-dimensional laser-diode arrays operating quasicontinuously. An analysis is reported of the influence of various structural parameters of a diode array on the thermal regime of a single laser. The temperature distributions along the cavity axis are investigated for different variants of mounting a crystal on a heat sink. It is found that the temperature drop along the cavity length in cw and quasi-cw laser diodes may exceed 20%. (lasers)
International Nuclear Information System (INIS)
Roussos, N.
1982-01-01
The main objective of this work is to create a neutronic calculations system for the SILOE-SILOETTE reactors, adaptable to other types of plate reactors. The author presents the methodology and the development of the APOLLO 1D (99 gr.) calculations for the creation of cross sections libraries. After a recall of the Discrete Ordinate Method (DOT), the method accuracy is studied in order to optimize the spatial discretization of the calculations; calculations of DOT 3.5 and of SILOETTE core are conducted and their convergence and costs are examined. DOT calculations of SILOETTE and experimental tests results are then compared [fr
Directory of Open Access Journals (Sweden)
Guodong Liu
2013-01-01
Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.
Calculation of two-dimensional thermal transients by the method of finite elements
International Nuclear Information System (INIS)
Fontoura Rodrigues, J.L.A. da.
1980-08-01
The unsteady linear heat conduction analysis throught anisotropic and/or heterogeneous matter, in either two-dimensional fields with any kind of geometry or three-dimensional fields with axial symmetry is presented. The boundary conditions and the internal heat generation are supposed time - independent. The solution is obtained by modal analysis employing the finite element method under Galerkin formulation. Optionally, it can be used with a reduced resolution method called Stoker Economizing Method wich allows a decrease on the program processing costs. (Author) [pt
Quantum Fidelity and Thermal Phase Transitions in a Two-Dimensional Spin System
International Nuclear Information System (INIS)
Wang Bo; Kou Su-Peng; Huang Hai-Lin; Sun Zhao-Yu
2012-01-01
We investigate the ability of quantum fidelity in detecting the classical phase transitions (CPTs) in a two-dimensional Heisenberg—Ising mixed spin model, which has a very rich phase diagram and is exactly soluble. For a two-site subsystem of the model, the reduced fidelity (including the operator fidelity and the fidelity susceptibility) at finite temperatures is calculated, and it is found that an extreme value presents at the critical temperature, thus shows a signal for the CPTs. In some parameter region, the signal becomes blurred. We propose to use the 'normalized fidelity susceptibility' to solve this problem
Development of an Advanced Two-Dimensional Thermal Model for Large size Lithium-ion Pouch Cells
International Nuclear Information System (INIS)
Samba, Ahmadou; Omar, Noshin; Gualous, Hamid; Firouz, Youssef; Van den Bossche, Peter; Van Mierlo, Joeri; Boubekeur, Tala Ighil
2014-01-01
In this work, a LiFePO4/graphite lithium-ion pouch cell with a rated capacity of 45Ah has been used and a two dimensional thermal model is developed to predict the cell temperature distribution over the surface of the battery, this model requires less input parameters and still has high accuracy. The used input parameters are the heat generation and thermal properties. The ANSYS FLUENT software has been used to solve the models. In addition, a new estimation tool has been developed for estimation of the thermal model parameters. Furthermore, the thermal behavior of the proposed battery has been investigated at different environmental conditions as well as during the abuse conditions. Thermal runaway is investigated in depth by the model
A two-dimensional model for transients calculations with phase changes in sodium cooled reactors
International Nuclear Information System (INIS)
Granziera, M.R.
1981-01-01
A computer code (NATOF2D) for the numerical simulation of situations where the radial non-uniformity in the sodium flow is an important factor, was developed. This computer code uses the two-fluid model, in which each phase is described by a complete set of mass conservation equations, energy equations and momentum equations. The experiment SLSF-P3A realized in the Engineering Test Reactor, Idaho, during the period of july to september of 1977, was simulated. (E.G.) [pt
International Nuclear Information System (INIS)
1981-01-01
This statement sets down briefly the CEGB's views on the requirement for nuclear power and outlines current progress in the implementation of the CEGB's thermal reactor strategy. The programme is traced historically, together with statements of Government policy. The place of Magnox, AGR, SGHWR, PWR and fast breeder reactors is discussed. Advantages and problems associated with the various types are outlined. (U.K.)
International Nuclear Information System (INIS)
Chen, G.S.; Christenson, J.M.
1985-01-01
In this paper, the authors present some initial results from an investigation of the application of a locally one-dimensional (LOD) finite difference method to the solution of the two-dimensional, two-group reactor kinetics equations. Although the LOD method is relatively well known, it apparently has not been previously applied to the space-time kinetics equations. In this investigation, the LOD results were benchmarked against similar computational results (using the same computing environment, the same programming structure, and the same sample problems) obtained by the TWIGL program. For all of the problems considered, the LOD method provided accurate results in one-half to one-eight of the time required by the TWIGL program
Directory of Open Access Journals (Sweden)
Yanjuan Wang
2017-10-01
Full Text Available Abstract: In this paper, the endothermic methanol decomposition reaction is used to obtain syngas by transforming middle and low temperature solar energy into chemical energy. A two-dimensional multiphysics coupling model of a middle and low temperature of 150~300 °C solar receiver/reactor was developed, which couples momentum equation in porous catalyst bed, the governing mass conservation with chemical reaction, and energy conservation incorporating conduction/convection/radiation heat transfer. The complex thermochemical conversion process of the middle and low temperature solar receiver/reactor (MLTSRR system was analyzed. The numerical finite element method (FEM model was validated by comparing it with the experimental data and a good agreement was obtained, revealing that the numerical FEM model is reliable. The characteristics of chemical reaction, coupled heat transfer, the components of reaction products, and the temperature fields in the receiver/reactor were also revealed and discussed. The effects of the annulus vacuum space and the glass tube on the performance of the solar receiver/reactor were further studied. It was revealed that when the direct normal irradiation increases from 200 W/m2 to 800 W/m2, the theoretical efficiency of solar energy transformed into chemical energy can reach 0.14–0.75. When the methanol feeding rate is 13 kg/h, the solar flux increases from 500 W/m2 to 1000 W/m2, methanol conversion can fall by 6.8–8.9% with air in the annulus, and methanol conversion can decrease by 21.8–28.9% when the glass is removed from the receiver/reactor.
Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests
International Nuclear Information System (INIS)
Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi
1987-01-01
Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)
John F. Hunt; Hongmei Gu
2006-01-01
The anisotropy of wood complicates solution of heat and mass transfer problems that require analyses be based on fundamental material properties of the wood structure. Most heat transfer models use average thermal properties across either the radial or tangential direction and do not differentiate the effects of cellular alignment, earlywood/latewood differences, or...
DEFF Research Database (Denmark)
Cui, Xiaoti; Kær, Søren Knudsen
2018-01-01
Monolithic catalysts have received increasing attention for application in the small-scale steam methane reforming process. The radial heat transfer behaviors of monolith reformers were analyzed by two-dimensional computational fluid dynamic (CFD) modeling. A parameter study was conducted...... by a large number of simulations focusing on the thermal conductivity of the monolith substrate, washcoat layer, wall gap, radiation heat transfer and the geometric parameters (cell density, porosity and diameter of monolith). The effective radial thermal conductivity of the monolith structure, kr......,eff, showed good agreement with predictions made by the pseudo-continuous symmetric model. This influence of the radiation heat transfer is low for highly conductive monoliths. A simplified model has been developed to evaluate the importance of radiation for monolithic reformers under different conditions...
International Nuclear Information System (INIS)
Wang Bo; Zhao Qinghe; Liu Lili; Gao Changyou; Han Kun; Zhang Junhu; Xiang Zheng; Yang Bai
2006-01-01
A novel and versatile soft lithography method, i.e. thermal pressing method has been established to create colloid arrays by using multilevel inks. Patterned poly(dimethylsiloxane) stamp containing silicone dioxide microparticles was pressed into a polycaprolactone (PCL) film at the temperature around the T m of PCL. Subsequent removal of the colloids left cavity arrays. By initially incorporating chitosan, albumin or CdTe quantum dots into the silicone dioxide microparticles, removal of the ordered SiO 2 microspheres would then release these substances which were stably embedded into the PCL matrices or suspended in the interiors of the cellular structures. By coating the SiO 2 microspheres with multilayers previously, thin covers on the cellular structures could be obtained after removal of the templates
Influence of Nanopore Shapes on Thermal Conductivity of Two-Dimensional Nanoporous Material.
Huang, Cong-Liang; Huang, Zun; Lin, Zi-Zhen; Feng, Yan-Hui; Zhang, Xin-Xin; Wang, Ge
2016-12-01
The influence of nanopore shapes on the electronic thermal conductivity (ETC) was studied in this paper. It turns out that with same porosity, the ETC will be quite different for different nanopore shapes, caused by the different channel width for different nanopore shapes. With same channel width, the influence of different nanopore shapes can be approximately omitted if the nanopore is small enough (smaller than 0.5 times EMFP in this paper). The ETC anisotropy was discovered for triangle nanopores at a large porosity with a large nanopore size, while there is a similar ETC for small pore size. It confirmed that the structure difference for small pore size may not be seen by electrons in their moving.
Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests
International Nuclear Information System (INIS)
Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi
1985-07-01
Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)
International Nuclear Information System (INIS)
Han, S.; Lapointe, J.; Lukens, J.E.
1992-01-01
The thermally induced escape rate of a particle trapped in a two-dimensional (2D) potential well has been investigated through experiment and numerical simulations. The measurements were performed on a special type of superconducting quantum interference device (SQUID) which has 2 degrees of freedom. The energies associated with the motion perpendicular to (transverse) and along (longitudinal) the escape direction are quite different: the ratio between the transverse and longitudinal small oscillation frequencies is ω t /ω l ∼7. The SQUID's parameters, which were used to determine the potential shape and energy scales were all independently determined. All data were obtained under conditions for which the 2D thermal activation (TA) model is expected to be valid. The results were found in good agreement with the theoretical prediction. The measured thermal activation energy is found to be the same as the barrier height calculated from the independently determined potential parameters. No evidence of apparent potential barrier enhancement recently reported in a similar system was found. In addition, the results of our numerical simulations suggest that the region in which the 2D thermal activation model is applicable may be extended to barriers as low as ΔU∼k BT
International Nuclear Information System (INIS)
Cao, Duc; Moses, Gregory; Delettrez, Jacques
2015-01-01
An implicit, non-local thermal conduction algorithm based on the algorithm developed by Schurtz, Nicolai, and Busquet (SNB) [Schurtz et al., Phys. Plasmas 7, 4238 (2000)] for non-local electron transport is presented and has been implemented in the radiation-hydrodynamics code DRACO. To study the model's effect on DRACO's predictive capability, simulations of shot 60 303 from OMEGA are completed using the iSNB model, and the computed shock speed vs. time is compared to experiment. Temperature outputs from the iSNB model are compared with the non-local transport model of Goncharov et al. [Phys. Plasmas 13, 012702 (2006)]. Effects on adiabat are also examined in a polar drive surrogate simulation. Results show that the iSNB model is not only capable of flux-limitation but also preheat prediction while remaining numerically robust and sacrificing little computational speed. Additionally, the results provide strong incentive to further modify key parameters within the SNB theory, namely, the newly introduced non-local mean free path. This research was supported by the Laboratory for Laser Energetics of the University of Rochester
Energy Technology Data Exchange (ETDEWEB)
Cao, Duc; Moses, Gregory [University of Wisconsin—Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States); Delettrez, Jacques [Laboratory for Laser Energetics of the University of Rochester, 250 East River Road, Rochester, New York 14623 (United States)
2015-08-15
An implicit, non-local thermal conduction algorithm based on the algorithm developed by Schurtz, Nicolai, and Busquet (SNB) [Schurtz et al., Phys. Plasmas 7, 4238 (2000)] for non-local electron transport is presented and has been implemented in the radiation-hydrodynamics code DRACO. To study the model's effect on DRACO's predictive capability, simulations of shot 60 303 from OMEGA are completed using the iSNB model, and the computed shock speed vs. time is compared to experiment. Temperature outputs from the iSNB model are compared with the non-local transport model of Goncharov et al. [Phys. Plasmas 13, 012702 (2006)]. Effects on adiabat are also examined in a polar drive surrogate simulation. Results show that the iSNB model is not only capable of flux-limitation but also preheat prediction while remaining numerically robust and sacrificing little computational speed. Additionally, the results provide strong incentive to further modify key parameters within the SNB theory, namely, the newly introduced non-local mean free path. This research was supported by the Laboratory for Laser Energetics of the University of Rochester.
Cao, Duc; Moses, Gregory; Delettrez, Jacques
2015-08-01
An implicit, non-local thermal conduction algorithm based on the algorithm developed by Schurtz, Nicolai, and Busquet (SNB) [Schurtz et al., Phys. Plasmas 7, 4238 (2000)] for non-local electron transport is presented and has been implemented in the radiation-hydrodynamics code DRACO. To study the model's effect on DRACO's predictive capability, simulations of shot 60 303 from OMEGA are completed using the iSNB model, and the computed shock speed vs. time is compared to experiment. Temperature outputs from the iSNB model are compared with the non-local transport model of Goncharov et al. [Phys. Plasmas 13, 012702 (2006)]. Effects on adiabat are also examined in a polar drive surrogate simulation. Results show that the iSNB model is not only capable of flux-limitation but also preheat prediction while remaining numerically robust and sacrificing little computational speed. Additionally, the results provide strong incentive to further modify key parameters within the SNB theory, namely, the newly introduced non-local mean free path. This research was supported by the Laboratory for Laser Energetics of the University of Rochester.
International Nuclear Information System (INIS)
1980-06-01
Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods
Energy Technology Data Exchange (ETDEWEB)
1980-06-01
Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.
International Nuclear Information System (INIS)
Nieves, Jose F.
2010-01-01
We apply the thermal field theory methods to study the propagation of photons in a plasma layer, that is a plasma in which the electrons are confined to a two-dimensional plane sheet. We calculate the photon self-energy and determine the appropriate expression for the photon propagator in such a medium, from which the properties of the propagating modes are obtained. The formulas for the photon dispersion relations and polarization vectors are derived explicitly in some detail for some simple cases of the thermal distributions of the charged particle gas, and appropriate formulas that are applicable in more general situations are also given.
REACTOR GROUT THERMAL PROPERTIES
Energy Technology Data Exchange (ETDEWEB)
Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.
2011-01-28
Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.
Directory of Open Access Journals (Sweden)
Kh. Lotfy
2013-01-01
Full Text Available The theory of two-temperature generalized thermoelasticity based on the theory of Youssef is used to solve boundary value problems of two-dimensional half-space. The governing equations are solved using normal mode method under the purview of the Lord-Şhulman (LS and the classical dynamical coupled theory (CD. The general solution obtained is applied to a specific problem of a half-space subjected to one type of heating, the thermal shock type. We study the influence of rotation on the total deformation of thermoelastic half-space and the interaction with each other under the influence of two temperature theory. The material is homogeneous isotropic elastic half-space. The methodology applied here is use of the normal mode analysis techniques that are used to solve the resulting nondimensional coupled field equations for the two theories. Numerical results for the displacement components, force stresses, and temperature distribution are presented graphically and discussed. The conductive temperature, the dynamical temperature, the stress, and the strain distributions are shown graphically with some comparisons.
International Nuclear Information System (INIS)
Mahmoud, K.S.; Szatmary, Z.
2005-01-01
An iterative method was developed for the numerical solution of the coupled two-dimensional time dependent multigroup diffusion equation and delayed precursor equations. Both forward (Explicit) and backward (Implicit) schemes were used. The second scheme was found to be numerically stable, while the first scheme requires that Δt -10 sec. for stability. An example is given for the second method. (authors)
Energy Technology Data Exchange (ETDEWEB)
Boyarinov, V. F.; Kondrushin, A. E.; Fomichenko, P. A. [National Research Centre Kurchatov Institute, Kurchatov Sq. 1, Moscow (Russian Federation)
2013-07-01
Time-dependent equations of the Surface Harmonics Method (SHM) have been derived from the time-dependent neutron transport equation with explicit representation of delayed neutrons for solving the two-dimensional time-dependent problems. These equations have been realized in the SUHAM-TD code. The TWIGL benchmark problem has been used for verification of the SUHAM-TD code. The results of the study showed that computational costs required to achieve necessary accuracy of the solution can be an order of magnitude less than with the use of the conventional finite difference method (FDM). (authors)
Thorium utilisation in thermal reactors
International Nuclear Information System (INIS)
Balakrishnan, K.
1997-01-01
It is now more or less accepted that the best way to use thorium is in thermal reactors. This is due to the fact that U233 is a good material in the thermal spectrum. Studies of different thorium cycles in various reactor concepts had been carried out in the early days of nuclear power. After three decades of neglect, the world is once again looking at thorium with some interest. We in India have been studying thorium cycles in most of the existing thermal reactor concepts, with greater emphasis on heavy water reactors. In this paper, we report some of the work done in India on different thorium cycles in the Indian pressurized heavy water reactor (PHWR), and also give a description of the design of the advanced heavy water reactor (AHWR). (author). 1 ref., 2 tabs., 5 figs
Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor
Energy Technology Data Exchange (ETDEWEB)
Ilas, Germina [ORNL; Primm, Trent [ORNL
2011-05-01
An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.
Two-dimensional ferroelectrics
Energy Technology Data Exchange (ETDEWEB)
Blinov, L M; Fridkin, Vladimir M; Palto, Sergei P [A.V. Shubnikov Institute of Crystallography, Russian Academy of Sciences, Moscow, Russian Federaion (Russian Federation); Bune, A V; Dowben, P A; Ducharme, Stephen [Department of Physics and Astronomy, Behlen Laboratory of Physics, Center for Materials Research and Analysis, University of Nebraska-Linkoln, Linkoln, NE (United States)
2000-03-31
The investigation of the finite-size effect in ferroelectric crystals and films has been limited by the experimental conditions. The smallest demonstrated ferroelectric crystals had a diameter of {approx}200 A and the thinnest ferroelectric films were {approx}200 A thick, macroscopic sizes on an atomic scale. Langmuir-Blodgett deposition of films one monolayer at a time has produced high quality ferroelectric films as thin as 10 A, made from polyvinylidene fluoride and its copolymers. These ultrathin films permitted the ultimate investigation of finite-size effects on the atomic thickness scale. Langmuir-Blodgett films also revealed the fundamental two-dimensional character of ferroelectricity in these materials by demonstrating that there is no so-called critical thickness; films as thin as two monolayers (1 nm) are ferroelectric, with a transition temperature near that of the bulk material. The films exhibit all the main properties of ferroelectricity with a first-order ferroelectric-paraelectric phase transition: polarization hysteresis (switching); the jump in spontaneous polarization at the phase transition temperature; thermal hysteresis in the polarization; the increase in the transition temperature with applied field; double hysteresis above the phase transition temperature; and the existence of the ferroelectric critical point. The films also exhibit a new phase transition associated with the two-dimensional layers. (reviews of topical problems)
International thermal reactor development
International Nuclear Information System (INIS)
Zebroski, E.L.
1977-01-01
The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities
International Nuclear Information System (INIS)
Adamik, V.; Matejovic, P.
1989-01-01
The problems are discussed of nonstationary, nonlinear dynamics of the continuum. A survey is presented of calculation methods in the given area with emphasis on the area of impact problems. A description is presented of the explicit finite elements method and its application to two-dimensional Cartesian and cylindrical configurations. Using the method the explicit calculation code FINEDAN was written which was tested in a series of verification calculations for different configurations and different types of continuum. The main characteristics are presented of the code and of some, of its practical applications. Envisaged trends of the development of the code and its possible applications in the technology of nuclear reactors are given. (author). 9 figs., 4 tabs., 10 refs
What can recycling in thermal reactors accomplish?
International Nuclear Information System (INIS)
Piet, Steven J.; Matthern, Gretchen E.; Jacobson, Jacob J.
2007-01-01
Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives. (authors)
What can Recycling in Thermal Reactors Accomplish?
International Nuclear Information System (INIS)
Steven Piet; Gretchen E. Matthern; Jacob J. Jacobson
2007-01-01
Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives
International Nuclear Information System (INIS)
Anon.
1991-01-01
This chapter addresses the extension of previous work in one-dimensional (linear) error theory to two-dimensional error analysis. The topics of the chapter include the definition of two-dimensional error, the probability ellipse, the probability circle, elliptical (circular) error evaluation, the application to position accuracy, and the use of control systems (points) in measurements
Hongmei Gu; John F. Hunt
2007-01-01
The anisotropy of wood creates a complex problem for solving heat and mass transfer problems that require analyses be based on fundamental material properties of the wood structure. Most heat transfer models for softwood use average thermal properties across either the radial or tangential direction and do not differentiate the effects of cellular alignment or...
Neutronic reactor thermal shield
International Nuclear Information System (INIS)
Lowe, P.E.
1976-01-01
A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures
Energy Technology Data Exchange (ETDEWEB)
Slater, C.O.
1990-07-01
Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.
International Nuclear Information System (INIS)
1980-06-01
Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport
Energy Technology Data Exchange (ETDEWEB)
1980-06-01
Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.
International Nuclear Information System (INIS)
Lee, Joon Hyun; Son, Bong Jin
1997-01-01
Although discontinuously reinforced metal matrix composite(MMC) is one of the most promising materials for applications of aerospace, automotive industries, the thermal residual stresses developed in the MMC due to the mismatch in coefficients of thermal expansion between the matrix and the fiber under a temperature change has been pointed out as one of the serious problem in practical applications. There are very limited nondestructive techniques to measure the residual stress of composite materials. However, many difficulties have been reported in their applications. Therefore it is important to establish analytical model to evaluate the thermal residual stress of MMC for practical engineering application. In this study, an elastic model is developed to predict the average thermal residual stresses in the matrix and fiber of a misoriented short fiber composite. The thermal residual stresses are induced by the mismatch in the coefficient of the thermal expansion of the matrix and fiber when the composite is subjected to a uniform temperature change. The model considers two-dimensional in-plane fiber misorientation. The analytical formulation of the model is based on Eshelby's equivalent inclusion method and is unique in that it is able to account for interactions among fibers. This model is more general than past models to investigate the effect of parameters which might influence thermal residual stress in composites. The present model is to investigate the effects of fiber volume fraction, distribution type, distribution cut-off angle, and aspect ratio on thermal residual stress for in-plane fiber misorientation. Fiber volume fraction, aspect ratio, and distribution cut-off angle are shown to have more significant effects on the magnitude of the thermal residual stresses than fiber distribution type for in-plane misorientation
Energy Technology Data Exchange (ETDEWEB)
Chafi, Fatima Zohra; Halle, Stephane [Mechanical engineering department, Ecole de technologie superieure, Quebec university, 1100 rue Notre-Dame Ouest, Montreal, Quebec H3C 1K3 (Canada)
2011-02-15
This paper presents the results of a study that consists of estimating the temperature distribution and air flow movement in a model room with a numerical model based on the Euler equations. Numerical results obtained for two scenarios of ventilation and heating are compared with the predictions of a Navier-Stokes model, as well as with experimental results. A comparison of the local thermal comfort indices PMV and PPD obtained experimentally and numerically is also presented. Results show that the Euler model is capable of properly estimating the temperature distribution, the air movement and the comfort indices in the room. Furthermore, the use of Euler equations allows a reduction of computational time in the order of 30% compared to the Navier-Stokes modeling. (author)
International Nuclear Information System (INIS)
Echigo, R.; Hasegawa, S.; Kamiuto, K.
1975-01-01
An analytical procedure is presented for simultaneous convective and radiative heat transfer with a fully developed laminar flow in a pipe by taking account of the two-dimensional propagation of radiative transfer and also shows the numerical results on the temperature profiles and the heat-transfer characteristics. In order to solve the energy equation with two-dimensional radiative transfer the entire ranges of the temperature field have to be solved simultaneously both along the radial and flow directions. Moreover, the heat flux by thermal radiation emitted from the heating wall propagates upstream so that it is necessary to examine the temperature profiles of the flowing medium to a certain distance upstream from the entrance of the heating section. In this way in order to attempt to solve the governing equation numerically by a finite difference method the dimension of matrix becomes extremely large provided that a satisfactory validity of numerical calculation is required Consequently the band matrix method is used and the temperature profiles of the medium in both regions upstream and downstream from the entrance of the heating section are illustrated and the heat transfer results are discussed in some detail by comparing with those of the one-dimensional transfer of radiation.(auth)
Spin dynamics, electronic, and thermal transport properties of two-dimensional CrPS4 single crystal
Pei, Q. L.; Luo, X.; Lin, G. T.; Song, J. Y.; Hu, L.; Zou, Y. M.; Yu, L.; Tong, W.; Song, W. H.; Lu, W. J.; Sun, Y. P.
2016-01-01
2-Dimensional (2D) CrPS4 single crystals have been grown by the chemical vapor transport method. The crystallographic, magnetic, electronic, and thermal transport properties of the single crystals were investigated by the room-temperature X-ray diffraction, electrical resistivity ρ(T), specific heat CP(T), and the electronic spin response (ESR) measurements. CrPS4 crystals crystallize into a monoclinic structure. The electrical resistivity ρ(T) shows a semiconducting behavior with an energy gap Ea = 0.166 eV. The antiferromagnetic transition temperature is about TN = 36 K. The spin flipping induced by the applied magnetic field is observed along the c axis. The magnetic phase diagram of CrPS4 single crystal has been discussed. The extracted magnetic entropy at TN is about 10.8 J/mol K, which is consistent with the theoretical value R ln(2S + 1) for S = 3/2 of the Cr3+ ion. Based on the mean-field theory, the magnetic exchange constants J1 and Jc corresponding to the interactions of the intralayer and between layers are about 0.143 meV and -0.955 meV are obtained based on the fitting of the susceptibility above TN, which agree with the results obtained from the ESR measurements. With the help of the strain for tuning the magnetic properties, monolayer CrPS4 may be a promising candidate to explore 2D magnetic semiconductors.
Energy Technology Data Exchange (ETDEWEB)
Pei, Q. L.; Luo, X., E-mail: xluo@issp.ac.cn, E-mail: ypsun@issp.ac.cn; Lin, G. T.; Song, J. Y.; Hu, L.; Song, W. H.; Lu, W. J. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zou, Y. M.; Yu, L.; Tong, W. [High Magnetic Field Laboratory, Chinese Academy of Sciences, Hefei 230031 (China); Sun, Y. P., E-mail: xluo@issp.ac.cn, E-mail: ypsun@issp.ac.cn [High Magnetic Field Laboratory, Chinese Academy of Sciences, Hefei 230031 (China); Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Collaborative Innovation Center of Advanced Microstructures, Nanjing University, Nanjing 210093 (China)
2016-01-28
2-Dimensional (2D) CrPS{sub 4} single crystals have been grown by the chemical vapor transport method. The crystallographic, magnetic, electronic, and thermal transport properties of the single crystals were investigated by the room-temperature X-ray diffraction, electrical resistivity ρ(T), specific heat C{sub P}(T), and the electronic spin response (ESR) measurements. CrPS{sub 4} crystals crystallize into a monoclinic structure. The electrical resistivity ρ(T) shows a semiconducting behavior with an energy gap E{sub a} = 0.166 eV. The antiferromagnetic transition temperature is about T{sub N} = 36 K. The spin flipping induced by the applied magnetic field is observed along the c axis. The magnetic phase diagram of CrPS{sub 4} single crystal has been discussed. The extracted magnetic entropy at T{sub N} is about 10.8 J/mol K, which is consistent with the theoretical value R ln(2S + 1) for S = 3/2 of the Cr{sup 3+} ion. Based on the mean-field theory, the magnetic exchange constants J{sub 1} and J{sub c} corresponding to the interactions of the intralayer and between layers are about 0.143 meV and −0.955 meV are obtained based on the fitting of the susceptibility above T{sub N}, which agree with the results obtained from the ESR measurements. With the help of the strain for tuning the magnetic properties, monolayer CrPS{sub 4} may be a promising candidate to explore 2D magnetic semiconductors.
International Nuclear Information System (INIS)
Chunxi, L.; Xuemin, Y.
2004-01-01
The temporal stability equation of the two-dimensional traveling waves of evaporating or condensing liquid films falling down on an inclined wall is established based on the Prandtl boundary layer theory and complete boundary conditions. The model indicates that the wave velocity is related to the effects of evaporating, isothermal and condensing states, thermo-capillarity, Reynolds number, fluid property and inclined angle, and the effects of above factors are distinctly different under different Reynolds numbers. The theoretical studies show that evaporation process induces the wave velocity to increase slightly compared with the isothermal case, and condensation process induces the wave velocity to decrease slightly. Furthermore, the wave velocity decreases because of the effects of thermo-capillarity under evaporation and increases because of the effects of thermo-capillarity under condensation. The effects of thermal non-equilibrium conditions have relatively obvious effects under lower Reynolds numbers and little effects under higher Reynolds numbers
International Nuclear Information System (INIS)
Tahir, N.A.; Kim, V.; Lamour, E.; Lomonosov, I.V.; Piriz, A.R.; Rozet, J.P.; Stöhlker, Th.; Sultanov, V.; Vernhet, D.
2012-01-01
In this paper we report on two-dimensional numerical simulations of heating of a rotating, wheel shaped target impacted by the full intensity of the ion beam that will be delivered by the SPIRAL2 facility at Caen, France. The purpose of this work is to study heating of solid targets that will be used to strip the fast ions of SPIRAL2 to the required high charge state for the FISIC (Fast Ion–Slow Ion Collision) experiments. Strippers of aluminum with different emissivities and of carbon are exposed to high beam current of different ion species as oxygen, neon and argon. These studies show that carbon, due to its much higher sublimation temperature and much higher emissivity, is more favorable compared to aluminum. For the highest beam intensities, an aluminum stripper does not survive. However, problem of the induced thermal stresses and long term material fatigue needs to be investigated before a final conclusion can be drawn.
Institute of Scientific and Technical Information of China (English)
Xiang Xu; Hao Ge; Shuai Wang; Zhongling Dai; Younian Wang; Aimin Zhu
2009-01-01
A two-dimensional (2D) fluid model is presented to study the discharge of argon in a dual frequency capacitively coupled plasma (CCP) reactor. We are interested in the influence of low-frequency (LF) source parameters such as applied voltage amplitudes and low frequencies on the plasma characteristics. In this paper, the high frequency is set to 60 MHz with voltage 50 V. The simulations were carried out for low frequencies of 1, 2 and 6 MHz with LF voltage 100 V, and for LF voltages of 60, 90 and 120 V with low frequency 2 MHz. The results of 2D distributions of electric field and ion density, the ion flux impinging on the substrate and the ion energy on the powered electrode are shown. As the low frequency increases, two sources become from uncoupling to coupling, When two sources are uncoupling, the increase in LF has little impact on the plasma characteristics, but when two sources are coupling, the increase in LF decreases the uniformities of ion density and ion flux noticeably. It is also found that with the increase in LF voltage, the uniformities in the radial direction of ion density distribution and ion flux at the powered electrode decreases significantly, and the energy of ions bombarding on the powered electrode increases significantly.
Wang, Zhibiao; Wang, Xu; Pei, Wenxuan; Li, Sen; Sun, Suqin; Zhou, Qun; Chen, Jianbo
2018-03-01
Areca semen is a common herb used in traditional Chinese medicine, but alkaloids in this herb are categorized as Group I carcinogens by IARC. It has been proven that the stir-baking process can reduce alkaloids in Areca semen while keep the activity for promoting digestion. However, the changes of compositions other than alkaloids during the thermal processing are unclear. Understanding the thermal chemical transitions of Areca semen is necessary to explore the processing mechanisms and optimize the procedures. In this research, FTIR spectroscopy with a temperature-controlled ATR accessory is employed to study the heating process of Areca semen. Principal component analysis and two-dimensional correlation spectroscopy are used to interpret the spectra to reveal the chemical transitions of Areca semen in different temperature ranges. The loss of a few volatile compounds in the testa and sperm happens below 105 °C, while some esters in the sperm decreases above 105 °C. As the heating temperature is close to 210 °C, Areca semen begins to be scorched and the decomposition of many compounds can be observed. This research shows the potential of the temperature-resolved ATR-FTIR spectroscopy in exploring the chemical transitions of the thermal processing of herbal materials.
Liu, Yue; Booth, Jean-Paul; Chabert, Pascal
2018-02-01
A Cartesian-coordinate two-dimensional electrostatic particle-in-cell/Monte Carlo collision (PIC/MCC) plasma simulation code is presented, including a new treatment of charge balance at dielectric boundaries. It is used to simulate an Ar plasma in a symmetric radiofrequency capacitively-coupled parallel-plate reactor with a thick (3.5 cm) dielectric side-wall. The reactor size (12 cm electrode width, 2.5 cm electrode spacing) and frequency (15 MHz) are such that electromagnetic effects can be ignored. The dielectric side-wall effectively shields the plasma from the enhanced electric field at the powered-grounded electrode junction, which has previously been shown to produce locally enhanced plasma density (Dalvie et al 1993 Appl. Phys. Lett. 62 3207-9 Overzet and Hopkins 1993 Appl. Phys. Lett. 63 2484-6 Boeuf and Pitchford 1995 Phys. Rev. E 51 1376-90). Nevertheless, enhanced electron heating is observed in a region adjacent to the dielectric boundary, leading to maxima in ionization rate, plasma density and ion flux to the electrodes in this region, and not at the reactor centre as would otherwise be expected. The axially-integrated electron power deposition peaks closer to the dielectric edge than the electron density. The electron heating components are derived from the PIC/MCC simulations and show that this enhanced electron heating results from increased Ohmic heating in the axial direction as the electron density decreases towards the side-wall. We investigated the validity of different analytical formulas to estimate the Ohmic heating by comparing them to the PIC results. The widespread assumption that a time-averaged momentum transfer frequency, v m , can be used to estimate the momentum change can cause large errors, since it neglects both phase and amplitude information. Furthermore, the classical relationship between the total electron current and the electric field must be used with caution, particularly close to the dielectric edge where the (neglected
Institute of Scientific and Technical Information of China (English)
ZHANG; Renhua; WANG; Jinfeng; ZHU; Caiying; SUN; Xiaomin
2004-01-01
After having analyzed the requirement on the aerodynamic earth's surface roughness in two-dimensional distribution in the research field of interaction between land surface and atmosphere, this paper presents a new way to calculate the aerodynamic roughness using the earth's surface geometric roughness retrieved from SAR (Synthetic Aperture Radar) and TM thermal infrared image data. On the one hand, the SPM (Small Perturbation Model) was used as a theoretical SAR backscattering model to describe the relationship between the SAR backscattering coefficient and the earth's surface geometric roughness and its dielectric constant retrieved from the physical model between the soil thermal inertia and the soil surface moisture with the simultaneous TM thermal infrared image data and the ground microclimate data. On the basis of the SAR image matching with the TM image, the non-volume scattering surface geometric information was obtained from the SPM model at the TM image pixel scale, and the ground pixel surface's equivalent geometric roughness-height standard RMS (Root Mean Square) was achieved from the geometric information by the transformation of the typical topographic factors. The vegetation (wheat, tree) height retrieved from spectrum model was also transferred into its equivalent geometric roughness. A completely two-dimensional distribution map of the equivalent geometric roughness over the experimental area was produced by the data mosaic technique. On the other hand, according to the atmospheric eddy currents theory, the aerodynamic surface roughness was iterated out with the atmosphere stability correction method using the wind and the temperature profiles data measured at several typical fields such as bare soil field and vegetation field. After having analyzed the effect of surface equivalent geometric roughness together with dynamic and thermodynamic factors on the aerodynamic surface roughness within the working area, this paper first establishes a scale
Two-dimensional flexible nanoelectronics
Akinwande, Deji; Petrone, Nicholas; Hone, James
2014-12-01
2014/2015 represents the tenth anniversary of modern graphene research. Over this decade, graphene has proven to be attractive for thin-film transistors owing to its remarkable electronic, optical, mechanical and thermal properties. Even its major drawback--zero bandgap--has resulted in something positive: a resurgence of interest in two-dimensional semiconductors, such as dichalcogenides and buckled nanomaterials with sizeable bandgaps. With the discovery of hexagonal boron nitride as an ideal dielectric, the materials are now in place to advance integrated flexible nanoelectronics, which uniquely take advantage of the unmatched portfolio of properties of two-dimensional crystals, beyond the capability of conventional thin films for ubiquitous flexible systems.
International Nuclear Information System (INIS)
Colombo, V; Ghedini, E; Gherardi, M; Sanibondi, P; Shigeta, M
2012-01-01
Nano-particle synthesis by means of inductively coupled plasma torches is a material process of large technological interest. Numerous parameters are involved in the optimization of this process; hence the development of numerical models for the prediction of thermal and magneto-fluid dynamics fields, precursor powder trajectories and thermal history, as well as nano-particle formation and growth, is necessary for the up-scaling of these devices from laboratory batch production to an industrial continuous process. In this work, a two-dimensional (2D) discrete-type model (nodal model) for the analysis of nano-powder nucleation and growth is presented, taking into account convection, diffusion and turbulent effects on particle formation. Discrete-type models feature high precision and reveal a great deal of information useful for clarifying the nano-particle formation process. Using Si as the precursor material, 2D simulations of a nano-particle synthesis RF plasma apparatus with a reaction chamber are carried out. Good agreement is found when comparing results obtained with this model with those coming from a well-established nucleation-coupled moment method. Moreover, the extended amount of obtainable information that characterizes the nodal model is underlined. (paper)
Ntlhokwe, Gaalebalwe; Tredoux, Andreas G J; Górecki, Tadeusz; Edwards, Matthew; Vestner, Jochen; Muller, Magdalena; Erasmus, Lené; Joubert, Elizabeth; Christel Cronje, J; de Villiers, André
2017-07-01
The applicability of comprehensive two-dimensional gas chromatography (GC×GC) using a single-stage thermal modulator was explored for the analysis of honeybush tea (Cyclopia spp.) volatile compounds. Headspace solid phase micro-extraction (HS-SPME) was used in combination with GC×GC separation on a non-polar × polar column set with flame ionisation (FID) detection for the analysis of fermented Cyclopia maculata, Cyclopia subternata and Cyclopia genistoides tea infusions of a single harvest season. Method optimisation entailed evaluation of the effects of several experimental parameters on the performance of the modulator, the choice of columns in both dimensions, as well as the HS-SPME extraction fibre. Eighty-four volatile compounds were identified by co-injection of reference standards. Principal component analysis (PCA) showed clear differentiation between the species based on their volatile profiles. Due to the highly reproducible separations obtained using the single-stage thermal modulator, multivariate data analysis was simplified. The results demonstrate both the complexity of honeybush volatile profiles and the potential of GC×GC separation in combination with suitable data analysis techniques for the investigation of the relationship between sensory properties and volatile composition of these products. The developed method therefore offers a fast and inexpensive methodology for the profiling of honeybush tea volatiles. Graphical abstract Surface plot obtained for the GC×GC-FID analysis of honeybush tea volatiles.
Reactor pressure vessel thermal annealing
International Nuclear Information System (INIS)
Lee, A.D.
1997-01-01
The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the
Reactor Thermal Hydraulic Numerical Calculation And Modeling
International Nuclear Information System (INIS)
Duong Ngoc Hai; Dang The Ba
2008-01-01
In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)
Shepelev, V. V.; Inogamov, N. A.
2018-01-01
There are various geometrical variants of laser illumination and target design. Important direction of investigations is connected with tightly focused action (spot size may be less than micron) onto a thin metal film: thickness of a film is just few skin-layer depths. Duration of a pulse is τ L ˜ 0.1 ps. In these conditions energy absorbed in a skin layer first propagates normally to a surface: gradient ∂Te /∂x dominates, here and below x and y are normal and lateral directions. This process in 1-2 ps homogenizes electron temperature T e along thickness of a film. We consider conditions when a film or is supported by weakly conducting substrate, or is free standing. Therefore all absorbed energy is confined inside the film. At the next stage the internal energy begin to flow along the lateral direction—thus direction of energy expansion is changed from x to y because of the heat non-penetrating boundary condition imposed on the rear-side of the film. At the short two-temperature stage of lateral expansion the thermal conductivity κ is high. After that electron and ion temperatures equilibrates and later on the heat propagates with usual value of κ. Lateral expansion cools down the hot spot on long time scales and finally the molten spot recrystallizes. Two-dimensional approach allows us to consider all these stages from propagation in x direction (normal to a film) to propagation in y direction (along a film).
Kraft, Vadim; Grützke, Martin; Weber, Waldemar; Menzel, Jennifer; Wiemers-Meyer, Simon; Winter, Martin; Nowak, Sascha
2015-08-28
A two-dimensional ion chromatography (IC/IC) technique with heart-cutting mode for the separation of ionic organophosphates was developed. These analytes are generated during thermal degradation of three different commercially available Selectilyte™ lithium ion battery electrolytes. The composition of the investigated electrolytes is based on 1M lithium hexafluorophosphate (LiPF6) dissolved in ethylene carbonate/dimethyl carbonate (50:50wt%, LP30), ethylene carbonate/diethyl carbonate (50:50wt%, LP40) and ethylene carbonate/ethyl methyl carbonate (50:50wt%, LP50). The organophosphates were pre-separated from PF6(-) anion on the low capacity A Supp 4 column, which was eluted with a gradient step containing acetonitrile. The fraction containing analytes was retarded on a pre-concentration column and after that transferred to the high capacity columns, where the separation was performed isocratically. Different stationary phases and eluents were applied on the 2nd dimension for the investigation of retention times, whereas the highly promising results were obtained with a high capacitive A Supp 10 column. The organophosphates generated in LP30 and LP40 electrolytes could be separated by application of an aqueous NaOH eluent providing fast analysis time within 35min. For the separation of the organophosphates of LP50 electrolyte due to its complexity a NaOH eluent containing a mixture of methanol/H2O was necessary. In addition, the developed two dimensional IC method was hyphenated to an inductively coupled plasma mass spectrometer (ICP-MS) using aqueous NaOH without organic modifiers. This proof of principle measurement was carried out for future quantitative investigation regarding the concentration of the ionic organophosphates. Furthermore, the chemical stability of several ionic organophosphates in water and acetonitrile at room temperature over a period of 10h was investigated. In both solvents no decomposition of the investigated analytes was observed and
Cordero, Chiara; Rubiolo, Patrizia; Reichenbach, Stephen E; Carretta, Andrea; Cobelli, Luigi; Giardina, Matthew; Bicchi, Carlo
2017-01-13
The possibility to transfer methods from thermal to differential-flow modulated comprehensive two-dimensional gas chromatographic (GC×GC) platforms is of high interest to improve GC×GC flexibility and increase the compatibility of results from different platforms. The principles of method translation are here applied to an original method, developed for a loop-type thermal modulated GC×GC-MS/FID system, suitable for quali-quantitative screening of suspected fragrance allergens. The analysis conditions were translated to a reverse-injection differential flow modulated platform (GC×2GC-MS/FID) with a dual-parallel secondary column and dual detection. The experimental results, for a model mixture of suspected volatile allergens and for raw fragrance mixtures of different composition, confirmed the feasibility of translating methods by preserving 1 D elution order, as well as the relative alignment of resulting 2D peak patterns. A correct translation produced several benefits including an effective transfer of metadata (compound names, MS fragmentation pattern, response factors) by automatic template transformation and matching from the original/reference method to its translated counterpart. The correct translation provided: (a) 2D pattern repeatability, (b) MS fragmentation pattern reliability for identity confirmation, and (c) comparable response factors and quantitation accuracy within a concentration range of three orders of magnitude. The adoption of a narrow bore (i.e. 0.1mm d c ) first-dimension column to operate under close-to-optimal conditions with the differential-flow modulation GC×GC platform was also advantageous in halving the total analysis under the translated conditions. Copyright © 2016 Elsevier B.V. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
Yan, Dongming [School of Civil and Architectural Engineering, Zhejiang University, Hangzhou 310058 (China); Hou, Peipei; Liu, Chang [State Key Laboratory of Silicon Materials, School of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Chai, Wenxiang [College of Materials Science and Engineering, China Jiliang University, Hangzhou 310018 (China); Zheng, Xuerong [State Key Laboratory of Silicon Materials, School of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Zhang, Luodong [School of Civil and Architectural Engineering, Zhejiang University, Hangzhou 310058 (China); Zhi, Mingjia; Zhou, Chunmei [State Key Laboratory of Silicon Materials, School of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Liu, Yi, E-mail: liuyimse@zju.edu.cn [State Key Laboratory of Silicon Materials, School of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China)
2016-09-15
Two new quaternary thioarsenates(III) NaAg{sub 2}AsS{sub 3}·H{sub 2}O (1) and KAg{sub 2}AsS{sub 3} (2) with high yields have been successfully prepared through a facile surfactant-thermal method. It is interesting that 2 can only be obtained with the aid of ethanediamine (en), which indicates that weak basicity of solvent is beneficial to the growth of 2 compared with 1. Both 1 and 2 feature the similar two-dimensional (2D) layer structures. However, the distortion of the primary honeycomb-like nets in 2 is more severe than that of 1, which demonstrates that Na{sup +} and K{sup +} cations have different structure directing effects on these two thioarsenates(III). Both experimental and theoretical studies confirm 1 and 2 are semiconductors with band gaps in the visible region. Our success in preparing these two quaternary thioarsenates(III) proves that surfactant-thermal technique is a powerful yet facile synthetic method to explore new complex chalcogenides. - Graphical abstract: Two new quaternary thioarsenates(III) NaAg{sub 2}AsS{sub 3}·H{sub 2}O (1) and KAg{sub 2}AsS{sub 3} (2) with high yields have been successfully prepared through a facile surfactant-thermal method. X-ray single crystal diffraction analyses demonstrate that Na{sup +} and K{sup +} cations have different structure directing effects on these two thioarsenates(III). Both experimental and theoretical studies confirm 1 and 2 are semiconductors with band gaps in the visible region. Display Omitted - Highlights: • NaAg{sub 2}AsS{sub 3}⋅H{sub 2}O (1) and KAg{sub 2}AsS{sub 3} (2) were prepared through surfactant-thermal method. • Crystal structures show Na{sup ±} and K{sup ±} have different structure directing effects. • The weak basicity of solvent is benefit to the growth of 2 compared with 1. • Experimental and theoretical studies confirm 1 and 2 are semiconductors.
Osserman, Robert
2011-01-01
The basic component of several-variable calculus, two-dimensional calculus is vital to mastery of the broader field. This extensive treatment of the subject offers the advantage of a thorough integration of linear algebra and materials, which aids readers in the development of geometric intuition. An introductory chapter presents background information on vectors in the plane, plane curves, and functions of two variables. Subsequent chapters address differentiation, transformations, and integration. Each chapter concludes with problem sets, and answers to selected exercises appear at the end o
International Nuclear Information System (INIS)
Schroer, Bert; Freie Universitaet, Berlin
2005-02-01
It is not possible to compactly review the overwhelming literature on two-dimensional models in a meaningful way without a specific viewpoint; I have therefore tacitly added to the above title the words 'as theoretical laboratories for general quantum field theory'. I dedicate this contribution to the memory of J. A. Swieca with whom I have shared the passion of exploring 2-dimensional models for almost one decade. A shortened version of this article is intended as a contribution to the project 'Encyclopedia of mathematical physics' and comments, suggestions and critical remarks are welcome. (author)
Nuclear reactor vessel fuel thermal insulating barrier
Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.
2013-03-19
The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.
Two-dimensional turbulent convection
Mazzino, Andrea
2017-11-01
We present an overview of the most relevant, and sometimes contrasting, theoretical approaches to Rayleigh-Taylor and mean-gradient-forced Rayleigh-Bénard two-dimensional turbulence together with numerical and experimental evidences for their support. The main aim of this overview is to emphasize that, despite the different character of these two systems, especially in relation to their steadiness/unsteadiness, turbulent fluctuations are well described by the same scaling relationships originated from the Bolgiano balance. The latter states that inertial terms and buoyancy terms balance at small scales giving rise to an inverse kinetic energy cascade. The main difference with respect to the inverse energy cascade in hydrodynamic turbulence [R. H. Kraichnan, "Inertial ranges in two-dimensional turbulence," Phys. Fluids 10, 1417 (1967)] is that the rate of cascade of kinetic energy here is not constant along the inertial range of scales. Thanks to the absence of physical boundaries, the two systems here investigated turned out to be a natural physical realization of the Kraichnan scaling regime hitherto associated with the elusive "ultimate state of thermal convection" [R. H. Kraichnan, "Turbulent thermal convection at arbitrary Prandtl number," Phys. Fluids 5, 1374-1389 (1962)].
Thermal embrittlement of reactor vessel steels
International Nuclear Information System (INIS)
Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.
1995-01-01
As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels
Thermal-hydraulics of actinide burner reactors
International Nuclear Information System (INIS)
Takizuka, Takakazu; Mukaiyama, Takehiko; Takano, Hideki; Ogawa, Toru; Osakabe, Masahiro.
1989-07-01
As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)
Chang, G. S.; Lillo, M. A.
2009-08-01
The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y
International Nuclear Information System (INIS)
Schmid, J.
1985-11-01
A package of updated computer codes for velocity and temperature field calculations for a fast reactor fuel subassembly (or its part) by the finite element method is described. Isoparametric triangular elements of the second degree are used. (author)
Thermal properties of reactors and some instabilities
Energy Technology Data Exchange (ETDEWEB)
Hearfield, F.
1979-03-01
A discussion covers the thermal properties of adiabatic reactors and the failure of the reaction rate to increase with increasing temperature due to depletion of reagents, transition to mass transfer control, or reduction of adsorption at catalytic surfaces; non-adiabatic reactors and factors upsetting the balance between heat generation and removal and possibly causing a runaway reaction, including loss of agitation loop circulation, and cooling or heating media; multiple steady states, i.e. multiple balances between heat generation and removal, for a continuous stirred tank reactor and the conditions necessary for stability of a steady state; and the temperature distribution in a tubular reactor, including mechanisms for feedback of heat from downstream to upstream in the reactor, e.g. heat conduction and radiation from hot catalyst, or an added heat exchanger. Three case histories are presented in which reactants accumulated in the reactors and cooling was decreased, permitting the occurrence of violent runaway reactions.
EL-2 reactor: Thermal neutron flux distribution
International Nuclear Information System (INIS)
Rousseau, A.; Genthon, J.P.
1958-01-01
The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)
Development of demonstration advanced thermal reactor
Energy Technology Data Exchange (ETDEWEB)
Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige
1982-08-01
The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.
Development of demonstration advanced thermal reactor
International Nuclear Information System (INIS)
Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.
1982-01-01
The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)
Utilization of thorium in thermal reactors
International Nuclear Information System (INIS)
Srinivasan, K.R.; Nakra, A.N.
1978-01-01
Large deposits of thorium are found in India. 233 U produced by neutron capture in 232 Th is a more valuable fuel for thermal reactors than the plutonium that results from capture in 238 U. These two facts are the main reasons for the interest in utilizing thorium in power reactors. But natural thorium does not contain any fissile material and its capture cross section is nearly two and a half times that of 238 U. These have made the fuelling cost high. However, in certain conditions and certain types of reactors the costs are comparable with those using uranium fuel. The relative cost effectiveness of different fuels is discussed. Apart from long term interest, the short term interest of using thorium fuel in RAPP type reactors is also briefly described. Finally the reactor physics experiments using thorium fuel and their comparison with calculations are presented. (author)
International Nuclear Information System (INIS)
Simon-Weidner, J.
1975-05-01
The digital program TIMTEM calculates twodimensional, nonlinear temperature fields of reactor components of complex structure; inhomogeneity and anisotropy are taken into account. Systems consisting of different materials and therefore having different temperature- and/or time-dependent material characteristics are allowed. Various local, time- and/or temperature-dependent boundary conditions can be considered, too, which may be locally different from each other or can be interconnected. (orig.) [de
Choice of thermal reactor systems: a report
Energy Technology Data Exchange (ETDEWEB)
1977-09-01
This is a report by the UK National Nuclear Corporation published by the UK Secretary of State for Energy (Mr. Benn) on 29th July 1977. It is concerned with the advantages and disadvantages of three thermal reactor systems -the AGR (advanced gas cooled reactor), the PWR (pressurised water reactor), and the SGHWR (steam generating heavy water reactor). The object was to help in the future choice of a thermal system for the UK to cover the next 25 years. The matter of export potential is also considered. A programme of four stations of 1100 to 1300 MW each over six years starting from 1979 was assumed. It is emphasised that a decision must be taken now both about reactor systems and actual orders. Headings are as follows: Extract from conclusions reached; Summary of main features of assessment; General conclusions regarding the following - safety, security of the investment, operational characteristics, development and launching requirements, effect on industry, and capital and generation costs. It is stated that in order to make an overall judgement on reactor choice the technical, commercial and social issues involved must be weighed in conjunction with cost differentials.
Thermal baffle for fast-breeder reactor
International Nuclear Information System (INIS)
Rylatt, J.A.
1977-01-01
A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel. 3 claims, 2 figures
Thermal-hydraulic analysis of nuclear reactors
Zohuri, Bahman
2015-01-01
This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play. Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...
Calculation of the neutron parameters of fast thermal reactor
International Nuclear Information System (INIS)
Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.
1975-01-01
The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)
Maximum neutron flux in thermal reactors
International Nuclear Information System (INIS)
Strugar, P.V.
1968-12-01
Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples
International Nuclear Information System (INIS)
Kemaneci, Efe; Graef, Wouter; Rahimi, Sara; Van Dijk, Jan; Kroesen, Gerrit; Carbone, Emile; Jimenez-Diaz, Manuel
2015-01-01
A microwave-induced oxygen plasma is simulated using both stationary and time-resolved modelling strategies. The stationary model is spatially resolved and it is self-consistently coupled to the microwaves (Jimenez-Diaz et al 2012 J. Phys. D: Appl. Phys. 45 335204), whereas the time-resolved description is based on a global (volume-averaged) model (Kemaneci et al 2014 Plasma Sources Sci. Technol. 23 045002). We observe agreement of the global model data with several published measurements of microwave-induced oxygen plasmas in both continuous and modulated power inputs. Properties of the microwave plasma reactor are investigated and corresponding simulation data based on two distinct models shows agreement on the common parameters. The role of the square wave modulated power input is also investigated within the time-resolved description. (paper)
International Nuclear Information System (INIS)
Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.
1982-03-01
The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: halves@iprj.uerj.br, E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pos-Graduacao em Modelagem Computacional
2015-07-01
A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S{sub N}) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S{sub N} discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S{sub N} transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S{sub N} eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)
International Nuclear Information System (INIS)
Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C.
2015-01-01
A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S N ) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S N discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S N transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S N eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)
Thermal and flow design of helium-cooled reactors
International Nuclear Information System (INIS)
Melese, G.; Katz, R.
1984-01-01
This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors
Historical perspective of thermal reactor safety in light water reactors
International Nuclear Information System (INIS)
Levy, S.
1986-01-01
A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective
International Nuclear Information System (INIS)
Shamasundar, B.I.; Fehrenbach, M.E.
1981-05-01
The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations
Two-dimensional NMR spectrometry
International Nuclear Information System (INIS)
Farrar, T.C.
1987-01-01
This article is the second in a two-part series. In part one (ANALYTICAL CHEMISTRY, May 15) the authors discussed one-dimensional nuclear magnetic resonance (NMR) spectra and some relatively advanced nuclear spin gymnastics experiments that provide a capability for selective sensitivity enhancements. In this article and overview and some applications of two-dimensional NMR experiments are presented. These powerful experiments are important complements to the one-dimensional experiments. As in the more sophisticated one-dimensional experiments, the two-dimensional experiments involve three distinct time periods: a preparation period, t 0 ; an evolution period, t 1 ; and a detection period, t 2
Quasi-two-dimensional holography
International Nuclear Information System (INIS)
Kutzner, J.; Erhard, A.; Wuestenberg, H.; Zimpfer, J.
1980-01-01
The acoustical holography with numerical reconstruction by area scanning is memory- and time-intensive. With the experiences by the linear holography we tried to derive a scanning for the evaluating of the two-dimensional flaw-sizes. In most practical cases it is sufficient to determine the exact depth extension of a flaw, whereas the accuracy of the length extension is less critical. For this reason the applicability of the so-called quasi-two-dimensional holography is appropriate. The used sound field given by special probes is divergent in the inclined plane and light focussed in the perpendicular plane using cylindrical lenses. (orig.) [de
HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor
International Nuclear Information System (INIS)
Finch, D.R.
1965-01-01
1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions
International Nuclear Information System (INIS)
HEARD, F.J.
1999-01-01
A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels
Energy Technology Data Exchange (ETDEWEB)
HEARD, F.J.
1999-04-09
A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.
Thermal analysis of the fuel of a power reactor
International Nuclear Information System (INIS)
Casadei, Alberto Luiz
1970-01-01
This dissertation presents the main values of maximum temperature of the central fuel rod of a power reactor, numerically calculated considering one-dimensional and two-dimensional conduction. The maximum temperature obtained with two-dimensional conduction is slightly lesser than the obtained when considering one-dimensional regime. Also, there exist complementary information on the process convergence and the precision to be adopted when reaching a satisfactory solution. The dissertation also presents brief considerations on the economical effects when adopting small parameter variations of nuclear power plant. (author)
Two-dimensional metamaterial optics
International Nuclear Information System (INIS)
Smolyaninov, I I
2010-01-01
While three-dimensional photonic metamaterials are difficult to fabricate, many new concepts and ideas in the metamaterial optics can be realized in two spatial dimensions using planar optics of surface plasmon polaritons. In this paper we review recent progress in this direction. Two-dimensional photonic crystals, hyperbolic metamaterials, and plasmonic focusing devices are demonstrated and used in novel microscopy and waveguiding schemes
Optimized two-dimensional Sn transport (BISTRO)
International Nuclear Information System (INIS)
Palmiotti, G.; Salvatores, M.; Gho, C.
1990-01-01
This paper reports on an S n two-dimensional transport module developed for the French fast reactor code system CCRR to optimize algorithms in order to obtain the best performance in terms of computational time. A form of diffusion synthetic acceleration was adopted, and a special effort was made to solve the associated diffusion equation efficiently. The improvements in the algorithms, along with the use of an efficient programming language, led to a significant gain in computational time with respect to the DOT code
Czech Academy of Sciences Publication Activity Database
Ahadiat, G.; Tabatabaee, M.; Gholivand, K.; Zare, K.; Dušek, Michal; Kučeráková, Monika
2017-01-01
Roč. 16, č. 1 (2017), s. 7-16 ISSN 1024-1221 R&D Projects: GA ČR(CZ) GA15-12653S; GA MŠk LO1603 EU Projects: European Commission(XE) CZ.2.16/3.1.00/24510 Institutional support: RVO:68378271 Keywords : bismuth coordination polymer * tartrate ligand * thermal decomposition * alpha-Bi 2 O 3 nanoparticles Subject RIV: BM - Solid Matter Physics ; Magnetism OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.) Impact factor: 0.565, year: 2016
Directory of Open Access Journals (Sweden)
V. Rajesh
2014-08-01
Full Text Available The interaction of free convection with thermal radiation of a viscous incompressible unsteady flow past a vertical plate with ramped wall temperature and mass diffusion is presented here, taking into account the homogeneous chemical reaction of first order. The fluid is gray, absorbing-emitting but non-scattering medium and the Rosseland approximation is used to describe the radiative flux in the energy equation. The dimensionless governing equations are solved using an implicit finite-difference method of the Crank-Nicolson type, which is stable and convergent. The velocity profiles are compared with the available theoretical solution and are found to be in good agreement. Numerical results for the velocity, the temperature, the concentration, the local and average skin friction, the Nusselt number and Sherwood number are shown graphically. This work has wide application in chemical and power engineering and also in the study of vertical air flow into the atmosphere. The present results can be applied to an important class of flows in which the driving force for the flow is provided by combination of the thermal and chemical species diffusion effects.
Micromachined two dimensional resistor arrays for determination of gas parameters
van Baar, J.J.J.; Verwey, Willem B.; Dijkstra, Mindert; Dijkstra, Marcel; Wiegerink, Remco J.; Lammerink, Theodorus S.J.; Krijnen, Gijsbertus J.M.; Elwenspoek, Michael Curt
A resistive sensor array is presented for two dimensional temperature distribution measurements in a micromachined flow channel. This allows simultaneous measurement of flow velocity and fluid parameters, like thermal conductivity, diffusion coefficient and viscosity. More general advantages of
International Nuclear Information System (INIS)
Shimizu, T.
1983-01-01
SPOTBOW computer program has been developed for predicting detailed temperature and turbulent flow velocity fields around distorted fuel pins in LMFBR fuel assemblies, in which pin to pin and pin to wrapper tube contacts may occur. The present study started from the requirement of reactor core designers to evaluate local hot spot temperature due to the wire contact effect and the pin bowing effect on cladding temperature distribution. This code calculates for both unbaffled and wire-wrapped pin bundles. The Galerkin method and iterative procedure were used to solve the basic equations which govern the local heat and momentum transfer in turbulent fluid flow around the distorted pins. Comparisons have been made with cladding temperatures measured in normal and distorted pin bundle mockups to check the validity of this code. Predicted peak temperatures in the vicinity of wire contact point were somewhat higher than the measured values, and the shape of the peaks agreed well with measurement. The changes of cladding temperature due to the decrease of gap width between bowing pin and adjacent pin were predicted well
Thermal reactor benchmark tests on JENDL-2
International Nuclear Information System (INIS)
Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.
1983-11-01
A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)
THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.
1984-07-01
The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.
THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code
International Nuclear Information System (INIS)
Vondy, D.R.
1984-07-01
The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations
Thermal Hydraulic Tests for Reactor Core Safety
Energy Technology Data Exchange (ETDEWEB)
Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)
2007-06-15
The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.
maximum neutron flux at thermal nuclear reactors
International Nuclear Information System (INIS)
Strugar, P.
1968-10-01
Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr
Kojima, Kensuke; Moriguchi, Hiroyuki; Hattori, Akihiro; Kaneko, Tomoyuki; Yasuda, Kenji
2003-11-01
We have developed a new method that enables agar microstructures to be used to cultivate cells and that allows cell network patterns to be controlled. The method makes use of non-contact three-dimensional photo-thermal etching with a 1480 nm infrared focused laser beam, which is strongly absorbed by water and agar gel, to form the shapes of agar microstructures. It allows microstructures to be easily formed in an agar layer within a few minutes, with cell-culture holes formed by the spot heating of a 100 mW laser and tunnels by the tracing of a 100 microm s(-1), 40 mW laser. We cultivated rat cardiac myocytes in adjacent microstructures and observed synchronized beating in them 90 min after they had made physical contact. Our results indicate that the system can make and use microstructures for cell-network cultivation in a minimal amount of time without any expensive microfabrication facilities or complicated procedures.
Two-dimensional topological photonics
Khanikaev, Alexander B.; Shvets, Gennady
2017-12-01
Originating from the studies of two-dimensional condensed-matter states, the concept of topological order has recently been expanded to other fields of physics and engineering, particularly optics and photonics. Topological photonic structures have already overturned some of the traditional views on wave propagation and manipulation. The application of topological concepts to guided wave propagation has enabled novel photonic devices, such as reflection-free sharply bent waveguides, robust delay lines, spin-polarized switches and non-reciprocal devices. Discrete degrees of freedom, widely used in condensed-matter physics, such as spin and valley, are now entering the realm of photonics. In this Review, we summarize the latest advances in this highly dynamic field, with special emphasis on the experimental work on two-dimensional photonic topological structures.
Two-dimensional thermofield bosonization
International Nuclear Information System (INIS)
Amaral, R.L.P.G.; Belvedere, L.V.; Rothe, K.D.
2005-01-01
The main objective of this paper was to obtain an operator realization for the bosonization of fermions in 1 + 1 dimensions, at finite, non-zero temperature T. This is achieved in the framework of the real-time formalism of Thermofield Dynamics. Formally, the results parallel those of the T = 0 case. The well-known two-dimensional Fermion-Boson correspondences at zero temperature are shown to hold also at finite temperature. To emphasize the usefulness of the operator realization for handling a large class of two-dimensional quantum field-theoretic problems, we contrast this global approach with the cumbersome calculation of the fermion-current two-point function in the imaginary-time formalism and real-time formalisms. The calculations also illustrate the very different ways in which the transmutation from Fermi-Dirac to Bose-Einstein statistics is realized
Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics
International Nuclear Information System (INIS)
1991-01-01
Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)
Thermally-insulating layer for nuclear reactors
International Nuclear Information System (INIS)
1975-01-01
The thermally-insulating layer has been designed both for insulating surfaces within the core of a nuclear reactor and transmitting loads such as the core-weight. Said layer comprises a layer of bricks and a layer of tiles with smaller clearance between the tiles than between the bricks, the latter having a reduced cross-section against the tiles so as to be surrounded by relatively large interconnected ducts forming a continuous chamber behind the tile-layer in order to induce a substantial decreases in the transverse flow of the reactor-core coolant. The core preferably comprises hexagonal columns supported by rhomb-shaped plates, with channels distributed so as to mix the coolant of twelve columns. The plates are separated from support-tiles by means of pillars [fr
Source driven breeding thermal power reactors
International Nuclear Information System (INIS)
Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba
1978-03-01
Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)
Two-dimensional critical phenomena
International Nuclear Information System (INIS)
Saleur, H.
1987-09-01
Two dimensional critical systems are studied using transformation to free fields and conformal invariance methods. The relations between the two approaches are also studied. The analytical results obtained generally depend on universality hypotheses or on renormalization group trajectories which are not established rigorously, so numerical verifications, mainly using the transfer matrix approach, are presented. The exact determination of critical exponents; the partition functions of critical models on toruses; and results as the critical point is approached are discussed [fr
Two dimensional unstable scar statistics.
Energy Technology Data Exchange (ETDEWEB)
Warne, Larry Kevin; Jorgenson, Roy Eberhardt; Kotulski, Joseph Daniel; Lee, Kelvin S. H. (ITT Industries/AES Los Angeles, CA)
2006-12-01
This report examines the localization of time harmonic high frequency modal fields in two dimensional cavities along periodic paths between opposing sides of the cavity. The cases where these orbits lead to unstable localized modes are known as scars. This paper examines the enhancements for these unstable orbits when the opposing mirrors are both convex and concave. In the latter case the construction includes the treatment of interior foci.
International Nuclear Information System (INIS)
Silagadze, Z.K.
2007-01-01
Two-dimensional generalization of the original peak finding algorithm suggested earlier is given. The ideology of the algorithm emerged from the well-known quantum mechanical tunneling property which enables small bodies to penetrate through narrow potential barriers. We merge this 'quantum' ideology with the philosophy of Particle Swarm Optimization to get the global optimization algorithm which can be called Quantum Swarm Optimization. The functionality of the newborn algorithm is tested on some benchmark optimization problems
Physics of Plutonium Recycling in Thermal Reactors
International Nuclear Information System (INIS)
Kinchin, G.H.
1967-01-01
A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)
Physics of Plutonium Recycling in Thermal Reactors
Energy Technology Data Exchange (ETDEWEB)
Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)
1967-09-15
A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)
Thermal Energetic Reactor with High Reproduction of Fission Materials
Directory of Open Access Journals (Sweden)
Vladimir M. Kotov
2012-01-01
On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.
International Nuclear Information System (INIS)
Eraslan, A.H.
1975-02-01
A far-field mathematical model is presented for numerical simulation of short-time (within tidal cycle) transient, two-dimensional temperature distributions in large coastal and offshore regions resulting from the condenser cooling water discharges of power plants. The Eulerian FLIDE (fluid-in-discrete-element) formulation employs the integral forms of the conservation principles for mass and thermal energy in variable-sized discrete elements that span the specific flow region. The contributions of vertical variations of the velocity components and temperature are rigorously incorporated in the development of depth-averaged, two-dimensional energy transport fluxes by spatially integrating the conservation equations over the enclosure surfaces of the discrete elements. The general mathematical formulation considers completely arbitrary, transient oceanic flow conditions, which include periodic tidal, geostrophic, and wind-induced currents, as locally specified inputs to the model. The thermal impact of a hypothetical, multiunit generating station in a coastal region is analyzed where the oceanic flow conditions are assumed to be strictly periodic tidal currents within any appreciable net drift of sufficient duration to remove the heated effluent. The numerical simulation indicates that the periodic flow conditions cause considerable variations in the temperature distributions during the day and the tidal cycles, which result in severe recirculation and re-entrainment of the heated water between the intakes and the discharges of the different units. This leads to a gradual, long-term increase of the temperatures in the immediate vicinity of the discharge structures and also in the far-field zone. (U.S.)
Nuclear vapor thermal reactor propulsion technology
International Nuclear Information System (INIS)
Maya, I.; Diaz, N.J.; Dugan, E.T.; Watanabe, Y.; McClanahan, J.A.; Wen-Hsiung Tu; Carman, R.L.
1993-01-01
The conceptual design of a nuclear rocket based on the vapor core reactor is presented. The Nuclear Vapor Thermal Rocket (NVTR) offers the potential for a specific impulse of 1000 to 1200 s at thrust-to-weight ratios of 1 to 2. The design is based on NERVA geometry and systems with the solid fuel replaced by uranium tetrafluoride (UF 4 ) vapor. The closed-loop core does not rely on hydrodynamic confinement of the fuel. The hydrogen propellant is separated from the UF 4 fuel gas by graphite structure. The hydrogen is maintained at high pressure (∼100 atm), and exits the core at 3,100 K to 3,500 K. Zirconium carbide and hafnium carbide coatings are used to protect the hot graphite from the hydrogen. The core is surrounded by beryllium oxide reflector. The nuclear reactor core has been integrated into a 75 klb engine design using an expander cycle and dual turbopumps. The NVTR offers the potential for an incremental technology development pathway to high performance gas core reactors. Since the fuel is readily available, it also offers advantages in the initial cost of development, as it will not require major expenditures for fuel development
Thermal shielding device in LMFBR type reactors
International Nuclear Information System (INIS)
Nakamura, Hiroshi.
1985-01-01
Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)
Thermal-hydraulic interfacing code modules for CANDU reactors
Energy Technology Data Exchange (ETDEWEB)
Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others
1997-07-01
The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.
Thermal-hydraulic interfacing code modules for CANDU reactors
International Nuclear Information System (INIS)
Liu, W.S.; Gold, M.; Sills, H.
1997-01-01
The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis
Two dimensional infinite conformal symmetry
International Nuclear Information System (INIS)
Mohanta, N.N.; Tripathy, K.C.
1993-01-01
The invariant discontinuous (discrete) conformal transformation groups, namely the Kleinian and Fuchsian groups Gamma (with an arbitrary signature) of H (the Poincare upper half-plane l) and the unit disc Delta are explicitly constructed from the fundamental domain D. The Riemann surface with signatures of Gamma and conformally invariant automorphic forms (functions) with Peterson scalar product are discussed. The functor, where the category of complex Hilbert spaces spanned by the space of cusp forms constitutes the two dimensional conformal field theory. (Author) 7 refs
Two-dimensional liquid chromatography
DEFF Research Database (Denmark)
Græsbøll, Rune
-dimensional separation space. Optimization of gradients in online RP×RP is more difficult than in normal HPLC as a result of the increased number of parameters and their influence on each other. Modeling the coverage of the compounds across the two-dimensional chromatogram as a result of a change in gradients could...... be used for optimization purposes, and reduce the time spend on optimization. In this thesis (chapter 6), and manuscript B, a measure of the coverage of the compounds in the twodimensional separation space is defined. It is then shown that this measure can be modeled for changes in the gradient in both...
Two-dimensional capillary origami
Energy Technology Data Exchange (ETDEWEB)
Brubaker, N.D., E-mail: nbrubaker@math.arizona.edu; Lega, J., E-mail: lega@math.arizona.edu
2016-01-08
We describe a global approach to the problem of capillary origami that captures all unfolded equilibrium configurations in the two-dimensional setting where the drop is not required to fully wet the flexible plate. We provide bifurcation diagrams showing the level of encapsulation of each equilibrium configuration as a function of the volume of liquid that it contains, as well as plots representing the energy of each equilibrium branch. These diagrams indicate at what volume level the liquid drop ceases to be attached to the endpoints of the plate, which depends on the value of the contact angle. As in the case of pinned contact points, three different parameter regimes are identified, one of which predicts instantaneous encapsulation for small initial volumes of liquid. - Highlights: • Full solution set of the two-dimensional capillary origami problem. • Fluid does not necessarily wet the entire plate. • Global energy approach provides exact differential equations satisfied by minimizers. • Bifurcation diagrams highlight three different regimes. • Conditions for spontaneous encapsulation are identified.
Two-dimensional capillary origami
International Nuclear Information System (INIS)
Brubaker, N.D.; Lega, J.
2016-01-01
We describe a global approach to the problem of capillary origami that captures all unfolded equilibrium configurations in the two-dimensional setting where the drop is not required to fully wet the flexible plate. We provide bifurcation diagrams showing the level of encapsulation of each equilibrium configuration as a function of the volume of liquid that it contains, as well as plots representing the energy of each equilibrium branch. These diagrams indicate at what volume level the liquid drop ceases to be attached to the endpoints of the plate, which depends on the value of the contact angle. As in the case of pinned contact points, three different parameter regimes are identified, one of which predicts instantaneous encapsulation for small initial volumes of liquid. - Highlights: • Full solution set of the two-dimensional capillary origami problem. • Fluid does not necessarily wet the entire plate. • Global energy approach provides exact differential equations satisfied by minimizers. • Bifurcation diagrams highlight three different regimes. • Conditions for spontaneous encapsulation are identified.
Two dimensional solid state NMR
International Nuclear Information System (INIS)
Kentgens, A.P.M.
1987-01-01
This thesis illustrates, by discussing some existing and newly developed 2D solid state experiments, that two-dimensional NMR of solids is a useful and important extension of NMR techniques. Chapter 1 gives an overview of spin interactions and averaging techniques important in solid state NMR. As 2D NMR is already an established technique in solutions, only the basics of two dimensional NMR are presented in chapter 2, with an emphasis on the aspects important for solid spectra. The following chapters discuss the theoretical background and applications of specific 2D solid state experiments. An application of 2D-J resolved NMR, analogous to J-resolved spectroscopy in solutions, to natural rubber is given in chapter 3. In chapter 4 the anisotropic chemical shift is mapped out against the heteronuclear dipolar interaction to obtain information about the orientation of the shielding tensor in poly-(oxymethylene). Chapter 5 concentrates on the study of super-slow molecular motions in polymers using a variant of the 2D exchange experiment developed by us. Finally chapter 6 discusses a new experiment, 2D nutation NMR, which makes it possible to study the quadrupole interaction of half-integer spins. 230 refs.; 48 figs.; 8 tabs
Two-dimensional quantum repeaters
Wallnöfer, J.; Zwerger, M.; Muschik, C.; Sangouard, N.; Dür, W.
2016-11-01
The endeavor to develop quantum networks gave rise to a rapidly developing field with far-reaching applications such as secure communication and the realization of distributed computing tasks. This ultimately calls for the creation of flexible multiuser structures that allow for quantum communication between arbitrary pairs of parties in the network and facilitate also multiuser applications. To address this challenge, we propose a two-dimensional quantum repeater architecture to establish long-distance entanglement shared between multiple communication partners in the presence of channel noise and imperfect local control operations. The scheme is based on the creation of self-similar multiqubit entanglement structures at growing scale, where variants of entanglement swapping and multiparty entanglement purification are combined to create high-fidelity entangled states. We show how such networks can be implemented using trapped ions in cavities.
Thermal hydraulic reactor safety analyses and experiments
International Nuclear Information System (INIS)
Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.
1989-04-01
The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)
Resonance shielding in thermal reactor lattices
International Nuclear Information System (INIS)
Rothenstein, W.; Taviv, E.; Aminpour, M.
1982-01-01
The theoretical foundations of a new methodology for the accurate treatment of resonance absorption in thermal reactor lattice analysis are presented. This methodology is based on the solution of the point-energy transport equation in its integral or integro-differential form for a heterogeneous lattice using detailed resonance cross-section profiles. The methodology is applied to LWR benchmark analysis, with emphasis on temperature dependence of resonance absorption during fuel depletion, spatial and mutual self-shielding, integral parameter analysis and treatment of cluster geometry. The capabilities of the OZMA code, which implements the new methodology are discussed. These capabilities provide a means against which simpler and more rapid resonance absorption algorithms can be checked. (author)
Methods for thermal reactor lattice calculations
International Nuclear Information System (INIS)
Schneider, A.
1976-12-01
The American code HAMMER and the British code WIMS, for the analysis of thermal reactor lattices, have been investigated. The primary objective of this investigation was to identify the causes for the discrepancies that exist between the calculated and the experimentally determined reactivity of clean critical experiments. Three phases have been undertaken in the research: (a) Detailed comparison between the group cross-sections used by the codes; (b) Definition of the various approximations incorporated into the codes; (c) Comparison between the values of a variety of reaction rates calculated by the two codes. It was concluded that the main cause of discrepancy between calculations and experiments is due to data inaccuracies, while approximations introduced in solving the transport equation are of smaller importance
JENDL-3.3 thermal reactor benchmark test
International Nuclear Information System (INIS)
Akie, Hiroshi
2001-01-01
Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)
Study of two-dimensional interchange turbulence
International Nuclear Information System (INIS)
Sugama, Hideo; Wakatani, Masahiro.
1990-04-01
An eddy viscosity model describing enstrophy transfer in two-dimensional turbulence is presented. This model is similar to that of Canuto et al. and provides an equation for the energy spectral function F(k) as a function of the energy input rate to the system per unit wavenumber, γ s (k). In the enstrophy-transfer inertial range, F(k)∝ k -3 is predicted by the model. The eddy viscosity model is applied to the interchange turbulence of a plasma in shearless magnetic field. Numerical simulation of the two-dimensional interchange turbulence demonstrates that the energy spectrum in the high wavenumber region is well described by this model. The turbulent transport driven by the interchange turbulence is expressed in terms of the Nusselt number Nu, the Rayleigh number Ra and Prantl number Pr in the same manner as that of thermal convection problem. When we use the linear growth rate for γ s (k), our theoretical model predicts that Nu ∝ (Ra·Pr) 1/2 for a constant background pressure gradient and Nu ∝ (Ra·Pr) 1/3 for a self-consistent background pressure profile with the stress-free slip boundary conditions. The latter agrees with our numerical result showing Nu ∝ Ra 1/3 . (author)
International Nuclear Information System (INIS)
Lin, J.; Millis, A.J.
2011-01-01
We calculate the frequency-dependent longitudinal (σ xx ) and Hall (σ xy ) conductivities for two-dimensional metals with thermally disordered antiferromagnetism using a generalization of a theoretical model, involving a one-loop quasistatic fluctuation approximation, which was previously used to calculate the electron self-energy. The conductivities are calculated from the Kubo formula, with current vertex function treated in a conserving approximation satisfying the Ward identity. In order to obtain a finite dc limit, we introduce phenomenologically impurity scattering, characterized by a relaxation time τ. σ xx ((Omega)) satisfies the f-sum rule. For the infinitely peaked spin-correlation function, χ(q)∝(delta)(q-Q), we recover the expressions for the conductivities in the mean-field theory of the ordered state. When the spin-correlation length ζ is large but finite, both σ xx and σ xy show behaviors characteristic of the state with long-range order. The calculation runs into difficulty for (Omega) ∼ xx ((Omega)) and σ xy ((Omega)) are qualitatively consistent with data on electron-doped cuprates when (Omega) > 1/τ.
Agapiou, A; Zorba, E; Mikedi, K; McGregor, L; Spiliopoulou, C; Statheropoulos, M
2015-07-09
Field experiments were devised to mimic the entrapment conditions under the rubble of collapsed buildings aiming to investigate the evolution of volatile organic compounds (VOCs) during the early dead body decomposition stage. Three pig carcasses were placed inside concrete tunnels of a search and rescue (SAR) operational field terrain for simulating the entrapment environment after a building collapse. The experimental campaign employed both laboratory and on-site analytical methods running in parallel. The current work focuses only on the results of the laboratory method using thermal desorption coupled to comprehensive two-dimensional gas chromatography with time-of-flight mass spectrometry (TD-GC×GC-TOF MS). The flow-modulated TD-GC×GC-TOF MS provided enhanced separation of the VOC profile and served as a reference method for the evaluation of the on-site analytical methods in the current experimental campaign. Bespoke software was used to deconvolve the VOC profile to extract as much information as possible into peak lists. In total, 288 unique VOCs were identified (i.e., not found in blank samples). The majority were aliphatics (172), aromatics (25) and nitrogen compounds (19), followed by ketones (17), esters (13), alcohols (12), aldehydes (11), sulfur (9), miscellaneous (8) and acid compounds (2). The TD-GC×GC-TOF MS proved to be a sensitive and powerful system for resolving the chemical puzzle of above-ground "scent of death". Copyright © 2015 Elsevier B.V. All rights reserved.
Equilibrium: two-dimensional configurations
International Nuclear Information System (INIS)
Anon.
1987-01-01
In Chapter 6, the problem of toroidal force balance is addressed in the simplest, nontrivial two-dimensional geometry, that of an axisymmetric torus. A derivation is presented of the Grad-Shafranov equation, the basic equation describing axisymmetric toroidal equilibrium. The solutions to equations provide a complete description of ideal MHD equilibria: radial pressure balance, toroidal force balance, equilibrium Beta limits, rotational transform, shear, magnetic wall, etc. A wide number of configurations are accurately modeled by the Grad-Shafranov equation. Among them are all types of tokamaks, the spheromak, the reversed field pinch, and toroidal multipoles. An important aspect of the analysis is the use of asymptotic expansions, with an inverse aspect ratio serving as the expansion parameter. In addition, an equation similar to the Grad-Shafranov equation, but for helically symmetric equilibria, is presented. This equation represents the leading-order description low-Beta and high-Beta stellarators, heliacs, and the Elmo bumpy torus. The solutions all correspond to infinitely long straight helices. Bending such a configuration into a torus requires a full three-dimensional calculation and is discussed in Chapter 7
International Nuclear Information System (INIS)
Quan, Xu; Qiang, Tian
2009-01-01
This paper discusses the two-dimensional discrete monatomic Fermi–Pasta–Ulam lattice, by using the method of multiple-scale and the quasi-discreteness approach. By taking into account the interaction between the atoms in the lattice and their nearest neighbours, it obtains some classes of two-dimensional local models as follows: two-dimensional bright and dark discrete soliton trains, two-dimensional bright and dark line discrete breathers, and two-dimensional bright and dark discrete breather. (condensed matter: structure, thermal and mechanical properties)
Thermal aspects of a superconducting coil for fusion reactor
International Nuclear Information System (INIS)
Yeh, H.T.
1975-01-01
Computer models are used to simulate both localized and extensive thermal excursions in a large superconducting magnet for fusion reactor. Conditions for the failure of fusion magnet due to thermal excursion are delineated. Designs to protect the magnet against such thermal excursion are evaluated
Thermal performances of an insulating structure for a reactor vessel
International Nuclear Information System (INIS)
Aranovitch, E.; Crutzen, S.; Le Det, M.; Denis, R.
1974-12-01
This report describes the thermal and technological tests performed on a multilayer thermal insulation system for high temperature gas reactors. It includes the description of test facilities, global tests, interpretation of data, and technological tests. Results obtained make it possible to predetermine with a satisfactory precision thermal performances under various nominal conditions
Analysis of thermal fatigue events in light water reactors
Energy Technology Data Exchange (ETDEWEB)
Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)
2000-09-01
Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)
Benchmark tests for fast and thermal reactor applications
International Nuclear Information System (INIS)
Seki, Yuji
1984-01-01
Integral tests of JENDL-2 library for fast and thermal reactor applications are reviewed including relevant analyses of JUPITER experiments. Criticality and core center characteristics were tested with one-dimensional models for a total of 27 fast critical assemblies. More sofisticated problems such as reaction rate distributions, control rod worths and sodium void reactivities were tested using two-dimensional models for MOZART and ZPPR-3 assemblies. Main observations from the fast core benchmark tests are as follows. 1) The criticality is well predicted; the average C/E value is 0.999+-0.008 for uranium cores and 0.997+-0.005 for plutonium cores. 2) The calculation underpredicts the reaction rate ratio 239 Pusub(fis)/ 235 Usub(fis) by 3% and overpredicts 238 Usub(cap)/ 239 Pusub(fis) by 6%. The results are consistent with those of JUPITER analyses. 3) The reaction rate distributions in the cores of prototype size are well predicted within +-3%. In larger JUPITER cores, however, the C/E value increases with the radial distance from the core center up to 6% at the outer core edge. 4) The prediction of control rod worths is satisfactory; C/E values are within the range from 0.92 to 0.97 with no apparent dependence on 10 B enrichment and the number of control rods inserted. Spatial dependence of C/E is also observed in the JUPITER cores. 5) The sodium void reactivity is overpredicted by 30% to 50% to the positive side. 1) The criticality is well predicted, as is the same in the fast core tests; the average C/E is 0.997+-0.003. 2) The calculation overpredicts 238 Usub(fis)/ 235 Usub(fis) by 3% to 6%, which shows the same tendency as in the small and medium size fast assemblies. The 238 Usub(cap)/ 235 Usub(fis) ratio is well predicted in the thermal cores. The calculated reaction rate ratios of 232 Th deviate from the measurements by 10% to 15%. (author)
Thermal Energetic Reactor with High Reproduction of Fission Materials
International Nuclear Information System (INIS)
Kotov, V.M.
2012-01-01
Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.
Thermal stratification of sodium in the BN 600 reactor
International Nuclear Information System (INIS)
Obmelukhin, J.A.; Obukhov, P.I.; Rinejskij, A.A.; Sobolev, V.A.; Sherbakov, S.I.
1983-01-01
The signs of thermal stratification of sodium in the BN 600 reactor upper plenum revealed by the analysis of standard temperature sensors' readings are defined. The initial conditions for existence of different temperature sodium layers are given. Two approaches for realizing on a computer of equations describing sodium motion in the upper plenum of the reactor are presented. (author)
Equipment for thermal neutron flux measurements in reactor R2
Energy Technology Data Exchange (ETDEWEB)
Johansson, E; Nilsson, T; Claeson, S
1960-04-15
For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.
Radiation effects on two-dimensional materials
Energy Technology Data Exchange (ETDEWEB)
Walker, R.C. II; Robinson, J.A. [Department of Materials Science, Penn State, University Park, PA (United States); Center for Two-Dimensional Layered Materials, Penn State, University Park, PA (United States); Shi, T. [Department of Mechanical and Nuclear Engineering, Penn State, University Park, PA (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI (United States); Silva, E.C. [GlobalFoundries, Malta, NY (United States); Jovanovic, I. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI (United States)
2016-12-15
The effects of electromagnetic and particle irradiation on two-dimensional materials (2DMs) are discussed in this review. Radiation creates defects that impact the structure and electronic performance of materials. Determining the impact of these defects is important for developing 2DM-based devices for use in high-radiation environments, such as space or nuclear reactors. As such, most experimental studies have been focused on determining total ionizing dose damage to 2DMs and devices. Total dose experiments using X-rays, gamma rays, electrons, protons, and heavy ions are summarized in this review. We briefly discuss the possibility of investigating single event effects in 2DMs based on initial ion beam irradiation experiments and the development of 2DM-based integrated circuits. Additionally, beneficial uses of irradiation such as ion implantation to dope materials or electron-beam and helium-beam etching to shape materials have begun to be used on 2DMs and are reviewed as well. For non-ionizing radiation, such as low-energy photons, we review the literature on 2DM-based photo-detection from terahertz to UV. The majority of photo-detecting devices operate in the visible and UV range, and for this reason they are the focus of this review. However, we review the progress in developing 2DMs for detecting infrared and terahertz radiation. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)
Thermal neutron flux distribution in ET-RR-2 reactor thermal column
Directory of Open Access Journals (Sweden)
Imam Mahmoud M.
2002-01-01
Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.
The dissolver paradox as a coupled fast-thermal reactor
International Nuclear Information System (INIS)
Lutz, H.F.; Webb, P.S.
1993-05-01
The dissolver paradox is treated as coupled fast-thermal reactors. Each reactor is sub-critical but the coupling is sufficient to form a critical system. The practical importance of the system occurs when the fast system by itself is mass limited and the thermal system by itself is volume limited. Numerous 1D calculations have been made to calculate the neutron multiplication parameters of the separate fast and thermal systems that occur in the dissolver paradox. A model has been developed to describe the coupling between the systems. Monte Carlo calculations using the MCNP code have tested the model
Topology optimization of two-dimensional waveguides
DEFF Research Database (Denmark)
Jensen, Jakob Søndergaard; Sigmund, Ole
2003-01-01
In this work we use the method of topology optimization to design two-dimensional waveguides with low transmission loss.......In this work we use the method of topology optimization to design two-dimensional waveguides with low transmission loss....
Two-dimensional transport of tokamak plasmas
International Nuclear Information System (INIS)
Hirshman, S.P.; Jardin, S.C.
1979-01-01
A reduced set of two-fluid transport equations is obtained from the conservation equations describing the time evolution of the differential particle number, entropy, and magnetic fluxes in an axisymmetric toroidal plasma with nested magnetic surfaces. Expanding in the small ratio of perpendicular to parallel mobilities and thermal conductivities yields as solubility constraints one-dimensional equations for the surface-averaged thermodynamic variables and magnetic fluxes. Since Ohm's law E +u x B =R', where R' accounts for any nonideal effects, only determines the particle flow relative to the diffusing magnetic surfaces, it is necessary to solve a single two-dimensional generalized differential equation, (partial/partialt) delpsi. (delp - J x B) =0, to find the absolute velocity of a magnetic surface enclosing a fixed toroidal flux. This equation is linear but nonstandard in that it involves flux surface averages of the unknown velocity. Specification of R' and the cross-field ion and electron heat fluxes provides a closed system of equations. A time-dependent coordinate transformation is used to describe the diffusion of plasma quantities through magnetic surfaces of changing shape
Applying chemical engineering concepts to non-thermal plasma reactors
Pedro AFFONSO, NOBREGA; Alain, GAUNAND; Vandad, ROHANI; François, CAUNEAU; Laurent, FULCHERI
2018-06-01
Process scale-up remains a considerable challenge for environmental applications of non-thermal plasmas. Undersanding the impact of reactor hydrodynamics in the performance of the process is a key step to overcome this challenge. In this work, we apply chemical engineering concepts to analyse the impact that different non-thermal plasma reactor configurations and regimes, such as laminar or plug flow, may have on the reactor performance. We do this in the particular context of the removal of pollutants by non-thermal plasmas, for which a simplified model is available. We generalise this model to different reactor configurations and, under certain hypotheses, we show that a reactor in the laminar regime may have a behaviour significantly different from one in the plug flow regime, often assumed in the non-thermal plasma literature. On the other hand, we show that a packed-bed reactor behaves very similarly to one in the plug flow regime. Beyond those results, the reader will find in this work a quick introduction to chemical reaction engineering concepts.
Primary system thermal hydraulics of future Indian fast reactors
Energy Technology Data Exchange (ETDEWEB)
Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)
2015-12-01
Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.
Plant with nuclear reactor, in particular a thermal reactor
International Nuclear Information System (INIS)
Straub, H.
1988-01-01
The reactor core of the plant has tubular and vertically movable control rods moved by a flow of coolant under pressure. Each control rod surrounds a similarly tubular guide rod, stationary relative to the reactor core, leaving an annular slot-like space therebetween. The inside of each guide rod forms a first pressure chamber supplied with the coolant under pressure. The upper end of each control rod is closed and has a vertical shaft that extends into the inside of the guide rod and forms therewith a second annular slot-like space. At least one first restriction is provided in the first annular slot-like space and at least one second restriction is provided in the second annular slot-like space. A second pressure chamber is formed between both restrictions. The coolant supplied to the guide rod thus returns to the pressure vessel surrounding the reactor core through the second annular slot-like space, the second pressure chamber and the first annular slot-like space. Controlling means are provided, with which pressure thrusts can be generated if necessary in the coolant within the first pressure chamber. (author) 5 refs., 10 figs
Thermal shield support degradation in pressurized water reactors
International Nuclear Information System (INIS)
Sweeney, F.J.; Fry, D.N.
1986-01-01
Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs
Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'
International Nuclear Information System (INIS)
Novelli, A.
1982-01-01
The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)
Experimental study on two-dimensional film flow with local measurement methods
Energy Technology Data Exchange (ETDEWEB)
Yang, Jin-Hwa, E-mail: evo03@snu.ac.kr [Nuclear Thermal-Hydraulic Engineering Laboratory, Seoul National University, Gwanak 599, Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-600 (Korea, Republic of); Cho, Hyoung-Kyu [Nuclear Thermal-Hydraulic Engineering Laboratory, Seoul National University, Gwanak 599, Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Seok [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-600 (Korea, Republic of); Euh, Dong-Jin, E-mail: djeuh@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-600 (Korea, Republic of); Park, Goon-Cherl [Nuclear Thermal-Hydraulic Engineering Laboratory, Seoul National University, Gwanak 599, Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)
2015-12-01
Highlights: • An experimental study on the two-dimensional film flow with lateral air injection was performed. • The ultrasonic thickness gauge was used to measure the local liquid film thickness. • The depth-averaged PIV (Particle Image Velocimetry) method was applied to measure the local liquid film velocity. • The uncertainty of the depth-averaged PIV was quantified with a validation experiment. • Characteristics of two-dimensional film flow were classified following the four different flow patterns. - Abstract: In an accident condition of a nuclear reactor, multidimensional two-phase flows may occur in the reactor vessel downcomer and reactor core. Therefore, those have been regarded as important issues for an advanced thermal-hydraulic safety analysis. In particular, the multi-dimensional two-phase flow in the upper downcomer during the reflood phase of large break loss of coolant accident appears with an interaction between a downward liquid and a transverse gas flow, which determines the bypass flow rate of the emergency core coolant and subsequently, the reflood coolant flow rate. At present, some thermal-hydraulic analysis codes incorporate multidimensional modules for the nuclear reactor safety analysis. However, their prediction capability for the two-phase cross flow in the upper downcomer has not been validated sufficiently against experimental data based on local measurements. For this reason, an experimental study was carried out for the two-phase cross flow to clarify the hydraulic phenomenon and provide local measurement data for the validation of the computational tools. The experiment was performed in a 1/10 scale unfolded downcomer of Advanced Power Reactor 1400 (APR1400). Pitot tubes, a depth-averaged PIV method and ultrasonic thickness gauge were applied for local measurement of the air velocity, the liquid film velocity and the liquid film thickness, respectively. The uncertainty of the depth-averaged PIV method for the averaged
Experimental study on two-dimensional film flow with local measurement methods
International Nuclear Information System (INIS)
Yang, Jin-Hwa; Cho, Hyoung-Kyu; Kim, Seok; Euh, Dong-Jin; Park, Goon-Cherl
2015-01-01
Highlights: • An experimental study on the two-dimensional film flow with lateral air injection was performed. • The ultrasonic thickness gauge was used to measure the local liquid film thickness. • The depth-averaged PIV (Particle Image Velocimetry) method was applied to measure the local liquid film velocity. • The uncertainty of the depth-averaged PIV was quantified with a validation experiment. • Characteristics of two-dimensional film flow were classified following the four different flow patterns. - Abstract: In an accident condition of a nuclear reactor, multidimensional two-phase flows may occur in the reactor vessel downcomer and reactor core. Therefore, those have been regarded as important issues for an advanced thermal-hydraulic safety analysis. In particular, the multi-dimensional two-phase flow in the upper downcomer during the reflood phase of large break loss of coolant accident appears with an interaction between a downward liquid and a transverse gas flow, which determines the bypass flow rate of the emergency core coolant and subsequently, the reflood coolant flow rate. At present, some thermal-hydraulic analysis codes incorporate multidimensional modules for the nuclear reactor safety analysis. However, their prediction capability for the two-phase cross flow in the upper downcomer has not been validated sufficiently against experimental data based on local measurements. For this reason, an experimental study was carried out for the two-phase cross flow to clarify the hydraulic phenomenon and provide local measurement data for the validation of the computational tools. The experiment was performed in a 1/10 scale unfolded downcomer of Advanced Power Reactor 1400 (APR1400). Pitot tubes, a depth-averaged PIV method and ultrasonic thickness gauge were applied for local measurement of the air velocity, the liquid film velocity and the liquid film thickness, respectively. The uncertainty of the depth-averaged PIV method for the averaged
Thermal hydraulic simulation of the CANDU nuclear reactor
Energy Technology Data Exchange (ETDEWEB)
Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)
2017-07-01
The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)
Thorium utilization: conversion ratio and fuel needs in thermal reactors
International Nuclear Information System (INIS)
Oosterkamp, W.J.
1975-01-01
As a preparatory study for thorium utilization in thermal reactors a study has been made of the fuel comsumption in existing reactor types. A quantitative description is given of the influence of enrichment, burnup, amount of structural material, choise of coolant and control requirements on the convertion ratio. The enrichment is an important factor and a low fuel comsumption can be achieved by increasing the enrichment
Coupled fast-thermal system at the 'RB' nuclear reactor
International Nuclear Information System (INIS)
Pesic, M.
1987-04-01
The results of the analyses of the possibility of the coupled fast-thermal system (CFTS) design at the 'RB' nuclear reactor are shown. As the proof of the theoretical analyses the first stage CFTS-1 has been designed, realized, and tested. The excellent agreement between the results of the CFTS-1 studies and the theoretical predictions opens a straight way to the second, the final stage - realization of the designed CFST at the 'RB' nuclear reactor. (author)
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.)
Source driven breeding thermal power reactors
International Nuclear Information System (INIS)
Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba
1978-03-01
The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that the DT device. The efficienct of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed. (aurthor)
First wall thermal--mechanical analyses of the reference theta-pinch reactor
International Nuclear Information System (INIS)
Krakowski, R.A.; Hagenson, R.L.; Cort, G.E.
1977-01-01
The thermal-mechanical response of the Reference Theta-Pinch Reactor (RTPR) first wall was analyzed. The first wall problems anticipated for a pulsed, high-β fusion power plant can be ameliorated by either alterations in the physics operating point, materials reengineering, or blanket/first wall reconfiguration. Within the latter ''configuration'' scenario, a two-fold approach has been adopted for the thermal-mechanical portion of the RTPR first wall technology assessment. First, a number of new first wall configurations (bonded or unbonded laminated composites, all-ceramic structures, protective and/or sacrificial ''bumpers'') were considered. Second, a more quantitative failure criterion, based on the developing theories of fracture mechanics, was identified. For each first wall configuration, transient heat transfer and thermoelastic stress calculations have been made. Two-dimensional finite element structural analyses have been made for a variety of mechanical boundary conditions. Only the Al 2 O 3 /Nb - 1 Zr system has been considered. The results of this study indicated a wide range of design solutions to the pulsed thermal stress problem anticipated for the RTPR
Density dependence of reactor performance with thermal confinement scalings
International Nuclear Information System (INIS)
Stotler, D.P.
1992-03-01
Energy confinement scalings for the thermal component of the plasma published thus far have a different dependence on plasma density and input power than do scalings for the total plasma energy. With such thermal scalings, reactor performance (measured by Q, the ratio of the fusion power to the sum of the ohmic and auxiliary input powers) worsens with increasing density. This dependence is the opposite of that found using scalings based on the total plasma energy, indicating that reactor operation concepts may need to be altered if this density dependence is confirmed in future research
Piezoelectricity in Two-Dimensional Materials
Wu, Tao; Zhang, Hua
2015-01-01
Powering up 2D materials: Recent experimental studies confirmed the existence of piezoelectricity - the conversion of mechanical stress into electricity - in two-dimensional single-layer MoS2 nanosheets. The results represent a milestone towards
Construction of two-dimensional quantum chromodynamics
Energy Technology Data Exchange (ETDEWEB)
Klimek, S.; Kondracki, W.
1987-12-01
We present a sketch of the construction of the functional measure for the SU(2) quantum chromodynamics with one generation of fermions in two-dimensional space-time. The method is based on a detailed analysis of Wilson loops.
Development of Two-Dimensional NMR
Indian Academy of Sciences (India)
Home; Journals; Resonance – Journal of Science Education; Volume 20; Issue 11. Development of Two-Dimensional NMR: Strucure Determination of Biomolecules in Solution. Anil Kumar. General Article Volume 20 Issue 11 November 2015 pp 995-1002 ...
Phase transitions in two-dimensional systems
International Nuclear Information System (INIS)
Salinas, S.R.A.
1983-01-01
Some experiences are related using synchrotron radiation beams, to characterize solid-liquid (fusion) and commensurate solid-uncommensurate solid transitions in two-dimensional systems. Some ideas involved in the modern theories of two-dimensional fusion are shortly exposed. The systems treated consist of noble gases (Kr,Ar,Xe) adsorbed in the basal plane of graphite and thin films formed by some liquid crystal shells. (L.C.) [pt
A two-dimensional, semi-analytic expansion method for nodal calculations
International Nuclear Information System (INIS)
Palmtag, S.P.
1995-08-01
Most modern nodal methods used today are based upon the transverse integration procedure in which the multi-dimensional flux shape is integrated over the transverse directions in order to produce a set of coupled one-dimensional flux shapes. The one-dimensional flux shapes are then solved either analytically or by representing the flux shape by a finite polynomial expansion. While these methods have been verified for most light-water reactor applications, they have been found to have difficulty predicting the large thermal flux gradients near the interfaces of highly-enriched MOX fuel assemblies. A new method is presented here in which the neutron flux is represented by a non-seperable, two-dimensional, semi-analytic flux expansion. The main features of this method are (1) the leakage terms from the node are modeled explicitly and therefore, the transverse integration procedure is not used, (2) the corner point flux values for each node are directly edited from the solution method, and a corner-point interpolation is not needed in the flux reconstruction, (3) the thermal flux expansion contains hyperbolic terms representing analytic solutions to the thermal flux diffusion equation, and (4) the thermal flux expansion contains a thermal to fast flux ratio term which reduces the number of polynomial expansion functions needed to represent the thermal flux. This new nodal method has been incorporated into the computer code COLOR2G and has been used to solve a two-dimensional, two-group colorset problem containing uranium and highly-enriched MOX fuel assemblies. The results from this calculation are compared to the results found using a code based on the traditional transverse integration procedure
Thermal performance and efficiency of supercritical nuclear reactors
International Nuclear Information System (INIS)
Romney Duffey; Tracy Zhou; Hussam Khartabil
2009-01-01
The paper reviews the major advances and innovative aspects of the thermal performance of recent concepts for super-critical water-cooled nuclear reactors (SCWR). The concepts are based on the extensive experience in the thermal power industry with super and ultra-supercritical boilers and turbines. The challenges and goals of increased efficiency, reduced cost, enhanced safety and co-generation have been pursued over the last ten years, and have resulted both in viable concepts and a vibrant defined R and D effort. The supercritical concept has wide acceptance among industry, as it reflects standard engineering practices and current thermal plant technology that is being already deployed. The SCWR concept represents a continuous development of water-cooled reactor technology, which utilizes the best and latest advances made in the thermal power industry. (author)
Thermal calculations for water cooled research reactors
International Nuclear Information System (INIS)
Fabrega, S.
1979-01-01
The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics
On the thermal scattering law data for reactor lattice calculations
International Nuclear Information System (INIS)
Trkov, A.; Mattes, M.
2004-01-01
Thermal scattering law data for hydrogen bound in water, hydrogen bound in zirconium hydride and deuterium bound in heavy water have been re-evaluated. The influence of the thermal scattering law data on critical lattices has been studied with detailed Monte Carlo calculations and a summary of results is presented for a numerical benchmark and for the TRIGA reactor benchmark. Systematics for a large sequence of benchmarks analysed with the WIMS-D lattice code are also presented. (author)
Ductile fracture estimation of reactor pressure vessel under thermal shock
International Nuclear Information System (INIS)
Takahashi, Jun; Sakai, Shinsuke; Okamura, Hiroyuki
1990-01-01
This paper presents a new scheme for the estimation of unstable ductile fracture of a reactor pressure vessel under thermal shock conditions. First, it is shown that the bending moment applied to the cracked section can be evaluated by considering the plastic deformation of the cracked section and the thermal deformation of the shell. As the contribution of the local thermal stress to the J-value is negligible, the J-value under thermal shock can be easily evaluated by using fully plastic solutions for the cracked part. Next, the phenomena of ductile fracture under thermal shock are expressed on the load-versus-displacement diagram which enables us to grasp the transient phenomena visually. In addition, several parametrical surveys are performed on the above diagram concerning the variation of (1) thermal shock conditions, (2) initial crack length, and (3) J-resistance curve (i.e. embrittlement by neutron irradiation). (author)
Thermal insulation of high temperature reactors
International Nuclear Information System (INIS)
Cornille, Y.
1975-01-01
Operating conditions of HTR thermal insulation are given and heat insulators currently developed are described (fibers kept in position by metallic structures). For future applications and higher temperatures, research is directed towards solutions using ceramics or associating fibers and ceramics [fr
International Nuclear Information System (INIS)
Lassmann, K.
1978-01-01
The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given. (Auth.)
A powerful methodology for reactor vessel pressurized thermal shock analysis
International Nuclear Information System (INIS)
Boucau, J.; Mager, T.
1994-01-01
The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs
Thermal-hydraulic methods in fast reactor safety
International Nuclear Information System (INIS)
Weber, D.P.; Briggs, L.L.
1985-01-01
Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided
Neutron and thermal dynamics of a gaseous core fission reactor
International Nuclear Information System (INIS)
van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.
1989-01-01
In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles
A survey of thorium utilization in thermal power reactors
International Nuclear Information System (INIS)
Oosterkamp, W.J.
1974-01-01
The present status of thorium utilization in thermal reactors HTGR's, HWR's and LWR's has been reviewed. Physics considerations are made to obtain the optimum use of thorium. Existing information on reprocessing and refabrication is given together with the properties of thorium metal and thoria
Two-dimensional Simulations of Correlation Reflectometry in Fusion Plasmas
International Nuclear Information System (INIS)
Valeo, E.J.; Kramer, G.J.; Nazikian, R.
2001-01-01
A two-dimensional wave propagation code, developed specifically to simulate correlation reflectometry in large-scale fusion plasmas is described. The code makes use of separate computational methods in the vacuum, underdense and reflection regions of the plasma in order to obtain the high computational efficiency necessary for correlation analysis. Simulations of Tokamak Fusion Test Reactor (TFTR) plasma with internal transport barriers are presented and compared with one-dimensional full-wave simulations. It is shown that the two-dimensional simulations are remarkably similar to the results of the one-dimensional full-wave analysis for a wide range of turbulent correlation lengths. Implications for the interpretation of correlation reflectometer measurements in fusion plasma are discussed
The role of heater thermal response in reactor thermal limits during oscillartory two-phase flows
Energy Technology Data Exchange (ETDEWEB)
Ruggles, A.E.; Brown, N.W. [Univ. of Tennessee, Knoxville, TN (United States); Vasil`ev, A.D. [Nuclear Safety Institute, Moscow, (Russian Federation); Wendel, M.W. [Oak Ridge National Lab., TN (United States)
1995-09-01
Analytical and numerical investigations of critical heat flux (CHF) and reactor thermal limits are conducted for oscillatory two-phase flows often associated with natural circulation conditions. It is shown that the CHF and associated thermal limits depend on the amplitude of the flow oscillations, the period of the flow oscillations, and the thermal properties and dimensions of the heater. The value of the thermal limit can be much lower in unsteady flow situations than would be expected using time average flow conditions. It is also shown that the properties of the heater strongly influence the thermal limit value in unsteady flow situations, which is very important to the design of experiments to evaluate thermal limits for reactor fuel systems.
Design of tandem mirror reactors with thermal barriers
International Nuclear Information System (INIS)
Carlson, G.A.
1980-01-01
End-plug technologies for tandem mirror reactors include high-field superconducting magnets, neutral beam injectors, and gyrotrons for electron cyclotron resonant heating (ECRH). In addition to their normal use for sustenance of the end-plug plasmas, neutral beam injectors are used for ''pumping'' trapped ions from the thermal barrier regions by charge exchange. An extra function of the axially directed pump beams is the removal of thermalized alpha particles from the reactor. The principles of tandem mirror operation with thermal barriers will be demonstrated in the upgrade of the Tandem Mirror Experiment (TMX-U) in 1981 and the tandem configuration of the Mirror fusion Test Facility (MFTF-B) in 1984
Thermal hydraulics analysis of the Advanced High Temperature Reactor
Energy Technology Data Exchange (ETDEWEB)
Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)
2015-12-01
Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.
Two-dimensional nuclear magnetic resonance spectroscopy
International Nuclear Information System (INIS)
Bax, A.; Lerner, L.
1986-01-01
Great spectral simplification can be obtained by spreading the conventional one-dimensional nuclear magnetic resonance (NMR) spectrum in two independent frequency dimensions. This so-called two-dimensional NMR spectroscopy removes spectral overlap, facilitates spectral assignment, and provides a wealth of additional information. For example, conformational information related to interproton distances is available from resonance intensities in certain types of two-dimensional experiments. Another method generates 1 H NMR spectra of a preselected fragment of the molecule, suppressing resonances from other regions and greatly simplifying spectral appearance. Two-dimensional NMR spectroscopy can also be applied to the study of 13 C and 15 N, not only providing valuable connectivity information but also improving sensitivity of 13 C and 15 N detection by up to two orders of magnitude. 45 references, 10 figures
Determination of reactor thermal power using a more accurate method
International Nuclear Information System (INIS)
Papuga, J.; Madron, F.; Pliska, J.
2005-01-01
Reactor thermal power is an important operational parameter in many respects such as nuclear safety, reactor physics or evaluation of turbine thermal performance. Thermal power of a pressurized water reactor is determined on the basis of the steam generator thermal balance. The balance can be made in several variants differing from one another by the selection of different measuring circuits whose data are used in the balancing. In principle, no one such variant gives the true value of the thermal power. Among the variant values, the one nearest to the unknown true value of reactor thermal power is probably the value calculated with the lowest uncertainty. The determination of such uncertainty is not easy and its value can make even several percent, which has significant economic consequences. This paper presents the method of data reconciliation and its application to the data of the third of Dukovany NPP. The data reconciliation method allows to exploit all the information which process data contain. It is based on the statistical adjustment of the redundant data in such a way that the adjusted data obey generally valid laws of nature (e.g. conservation laws). Mass and energy balances based on the data not yet reconciled do not obey those laws because of measurement errors. For data reconciliation in Dukovany, a detailed model of mass and energy flows describing the 3rd unit from steam generators to alternator and condenser was set up. Laws of mass and energy conservation and phase equilibrium in water-steam systems are thus fulfilled. Moreover, the user can model momentum balances in pipelines and create other equations, which are respected during calculation. The data reconciliation is done regularly for hourly averages (Authors)
Two-dimensional x-ray diffraction
He, Bob B
2009-01-01
Written by one of the pioneers of 2D X-Ray Diffraction, this useful guide covers the fundamentals, experimental methods and applications of two-dimensional x-ray diffraction, including geometry convention, x-ray source and optics, two-dimensional detectors, diffraction data interpretation, and configurations for various applications, such as phase identification, texture, stress, microstructure analysis, crystallinity, thin film analysis and combinatorial screening. Experimental examples in materials research, pharmaceuticals, and forensics are also given. This presents a key resource to resea
Equivalence of two-dimensional gravities
International Nuclear Information System (INIS)
Mohammedi, N.
1990-01-01
The authors find the relationship between the Jackiw-Teitelboim model of two-dimensional gravity and the SL(2,R) induced gravity. These are shown to be related to a two-dimensional gauge theory obtained by dimensionally reducing the Chern-Simons action of the 2 + 1 dimensional gravity. The authors present an explicit solution to the equations of motion of the auxiliary field of the Jackiw-Teitelboim model in the light-cone gauge. A renormalization of the cosmological constant is also given
Transmutation of LWR waste actinides in thermal reactors
International Nuclear Information System (INIS)
Gorrell, T.C.
1979-01-01
Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles
Analytical simulation of two dimensional advection dispersion ...
African Journals Online (AJOL)
The study was designed to investigate the analytical simulation of two dimensional advection dispersion equation of contaminant transport. The steady state flow condition of the contaminant transport where inorganic contaminants in aqueous waste solutions are disposed of at the land surface where it would migrate ...
Analytical Simulation of Two Dimensional Advection Dispersion ...
African Journals Online (AJOL)
ADOWIE PERE
ABSTRACT: The study was designed to investigate the analytical simulation of two dimensional advection dispersion equation of contaminant transport. The steady state flow condition of the contaminant transport where inorganic contaminants in aqueous waste solutions are disposed of at the land surface where it would ...
Sums of two-dimensional spectral triples
DEFF Research Database (Denmark)
Christensen, Erik; Ivan, Cristina
2007-01-01
construct a sum of two dimensional modules which reflects some aspects of the topological dimensions of the compact metric space, but this will only give the metric back approximately. At the end we make an explicit computation of the last module for the unit interval in. The metric is recovered exactly...
Stability of two-dimensional vorticity filaments
International Nuclear Information System (INIS)
Elhmaidi, D.; Provenzale, A.; Lili, T.; Babiano, A.
2004-01-01
We discuss the results of a numerical study on the stability of two-dimensional vorticity filaments around a circular vortex. We illustrate how the stability of the filaments depends on the balance between the strain associated with the far field of the vortex and the local vorticity of the filament, and we discuss an empirical criterion for filament stability
Two-Dimensional Motions of Rockets
Kang, Yoonhwan; Bae, Saebyok
2007-01-01
We analyse the two-dimensional motions of the rockets for various types of rocket thrusts, the air friction and the gravitation by using a suitable representation of the rocket equation and the numerical calculation. The slope shapes of the rocket trajectories are discussed for the three types of rocket engines. Unlike the projectile motions, the…
Two-dimensional microstrip detector for neutrons
Energy Technology Data Exchange (ETDEWEB)
Oed, A [Institut Max von Laue - Paul Langevin (ILL), 38 - Grenoble (France)
1997-04-01
Because of their robust design, gas microstrip detectors, which were developed at ILL, can be assembled relatively quickly, provided the prefabricated components are available. At the beginning of 1996, orders were received for the construction of three two-dimensional neutron detectors. These detectors have been completed. The detectors are outlined below. (author). 2 refs.
Conformal invariance and two-dimensional physics
International Nuclear Information System (INIS)
Zuber, J.B.
1993-01-01
Actually, physicists and mathematicians are very interested in conformal invariance: geometric transformations which keep angles. This symmetry is very important for two-dimensional systems as phase transitions, string theory or node mathematics. In this article, the author presents the conformal invariance and explains its usefulness
Matching Two-dimensional Gel Electrophoresis' Spots
DEFF Research Database (Denmark)
Dos Anjos, António; AL-Tam, Faroq; Shahbazkia, Hamid Reza
2012-01-01
This paper describes an approach for matching Two-Dimensional Electrophoresis (2-DE) gels' spots, involving the use of image registration. The number of false positive matches produced by the proposed approach is small, when compared to academic and commercial state-of-the-art approaches. This ar...
Two-dimensional membranes in motion
Davidovikj, D.
2018-01-01
This thesis revolves around nanomechanical membranes made of suspended two - dimensional materials. Chapters 1-3 give an introduction to the field of 2D-based nanomechanical devices together with an overview of the underlying physics and the measurementtools used in subsequent chapters. The research
Extended Polymorphism of Two-Dimensional Material
Yoshida, Masaro; Ye, Jianting; Zhang, Yijin; Imai, Yasuhiko; Kimura, Shigeru; Fujiwara, Akihiko; Nishizaki, Terukazu; Kobayashi, Norio; Nakano, Masaki; Iwasa, Yoshihiro
When controlling electronic properties of bulk materials, we usually assume that the basic crystal structure is fixed. However, in two-dimensional (2D) materials, atomic structure or to functionalize their properties. Various polymorphs can exist in transition metal dichalcogenides (TMDCs) from
Piezoelectricity in Two-Dimensional Materials
Wu, Tao
2015-02-25
Powering up 2D materials: Recent experimental studies confirmed the existence of piezoelectricity - the conversion of mechanical stress into electricity - in two-dimensional single-layer MoS2 nanosheets. The results represent a milestone towards embedding low-dimensional materials into future disruptive technologies. © 2015 Wiley-VCH Verlag GmbH & Co. KGaA.
Boron nitride as two dimensional dielectric: Reliability and dielectric breakdown
Energy Technology Data Exchange (ETDEWEB)
Ji, Yanfeng; Pan, Chengbin; Hui, Fei; Shi, Yuanyuan; Lanza, Mario, E-mail: mlanza@suda.edu.cn [Institute of Functional Nano and Soft Materials, Collaborative Innovation Center of Suzhou Nano Science and Technology, Soochow University, 199 Ren-Ai Road, Suzhou 215123 (China); Zhang, Meiyun; Long, Shibing [Key Laboratory of Microelectronics Devices & Integrated Technology, Institute of Microelectronics, Chinese Academy of Sciences, Beijing 100029 (China); Lian, Xiaojuan; Miao, Feng [National Laboratory of Solid State Microstructures, School of Physics, Collaborative Innovation Center of Advanced Microstructures, Nanjing University, Nanjing 210093 (China); Larcher, Luca [DISMI, Università di Modena e Reggio Emilia, 42122 Reggio Emilia (Italy); Wu, Ernest [IBM Research Division, Essex Junction, Vermont 05452 (United States)
2016-01-04
Boron Nitride (BN) is a two dimensional insulator with excellent chemical, thermal, mechanical, and optical properties, which make it especially attractive for logic device applications. Nevertheless, its insulating properties and reliability as a dielectric material have never been analyzed in-depth. Here, we present the first thorough characterization of BN as dielectric film using nanoscale and device level experiments complementing with theoretical study. Our results reveal that BN is extremely stable against voltage stress, and it does not show the reliability problems related to conventional dielectrics like HfO{sub 2}, such as charge trapping and detrapping, stress induced leakage current, and untimely dielectric breakdown. Moreover, we observe a unique layer-by-layer dielectric breakdown, both at the nanoscale and device level. These findings may be of interest for many materials scientists and could open a new pathway towards two dimensional logic device applications.
Pellet bed reactor for nuclear thermal propelled vehicles
International Nuclear Information System (INIS)
El-Genk, M.; Morley, N.J.; Haloulakos, V.E.
1991-01-01
The Pellet Bed Reactor (PeBR) concept is capable of operating at a high power density of up to 3.0 kWt/cu cm and an exit hydrogen gas temperature of 3000 K. The nominal reactor thermal power is 1500 MW and the reactor core is 0.80 m in diameter and 1.3 m high. The nominal PeBR engine generates a thrust of approximately 315 kN at a specific impulse of 1000 s for a mission duration to Mars of 250 days requiring a total firing time of 170 minutes. Because of its low diameter-to-height ratio, PeBR has enough surface area for passive removal of the decay heat from the reactor core. The reactor is equipped with two independent shutdown mechanisms; 8-B4C safety rods and 26 BeO/B4C control drums; each system is capable of operating and scraming the reactor safely. Due to the absence of core internal support structures, the PeBR can be fueled and refueled in orbit using the vacuum of space. These unique features of the PeBR provide for safety during launch, simplicity of handling, deployment, and end-of-life disposal, and vehicle extended lifetime. 11 refs
Two-dimensional shielding benchmarks for iron at YAYOI, (1)
International Nuclear Information System (INIS)
Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.
The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)
International Nuclear Information System (INIS)
Kawahara, Toshio; Matsushita, Tadashi
1977-01-01
The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)
Modeling solid thermal explosion containment on reactor HNIW and HMX
International Nuclear Information System (INIS)
Lin, Chun-Ping; Chang, Chang-Ping; Chou, Yu-Chuan; Chu, Yung-Chuan; Shu, Chi-Min
2010-01-01
2,4,6,8,10,12-Hexanitro-2,4,6,8,10,12-hexaaza-isowurtzitane (HNIW), also known as CL-20 and octahydro-1,3,5,7-tetranitro-1,3,5,7-tetrazocine (HMX), are highly energetic materials which have been popular in national defense industries for years. This study established the models of thermal decomposition and thermal explosion hazard for HNIW and HMX. Differential scanning calorimetry (DSC) data were used for parameters determination of the thermokinetic models, and then these models were employed for simulation of thermal explosion in a 437 L barrel reactor and a 24 kg cubic box package. Experimental results indicating the best storage conditions to avoid any violent runaway reaction of HNIW and HMX were also discovered. This study also developed an efficient procedure regarding creation of thermokinetics and assessment of thermal hazards of HNIW and HMX that could be applied to ensure safe storage conditions.
Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Baek, W. P.; Song, C. H.; Kim, Y. S. and others
2005-02-15
The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.
Uranium-fuel thermal reactor benchmark testing of CENDL-3
International Nuclear Information System (INIS)
Liu Ping
2001-01-01
CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)
Thermal annealing of an embrittled reactor pressure vessel
International Nuclear Information System (INIS)
Mager, T.R.; Dragunov, Y.G.; Leitz, C.
1998-01-01
As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements
Requirements for thermal insulation on prestressed concrete reactor vessels
International Nuclear Information System (INIS)
Neylan, A.J.; Wistrom, J.D.
1979-01-01
During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)
Thermal-hydraulic modeling of porous bed reactors
International Nuclear Information System (INIS)
Araj, K.J.; Nourbakhsh, H.P.
1987-01-01
Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs
Thermal-hydraulic tests for reactor safety system
International Nuclear Information System (INIS)
Chun, Se Young; Chung, Moon Ki; Baek, Won Pil
2002-05-01
Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena
COOLOD, Steady-State Thermal Hydraulics of Research Reactors
International Nuclear Information System (INIS)
Kaminaga, Masanori
1997-01-01
1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow
An analytical method for neutron thermalization calculations in heterogenous reactors
Energy Technology Data Exchange (ETDEWEB)
Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)
1965-07-01
It is well known that the use of the diffusion approximation for stu
Numerical Modelling of Wood Gasification in Thermal Plasma Reactor
Czech Academy of Sciences Publication Activity Database
Hirka, Ivan; Živný, Oldřich; Hrabovský, Milan
2017-01-01
Roč. 37, č. 4 (2017), s. 947-965 ISSN 0272-4324 Institutional support: RVO:61389021 Keywords : Plasma modelling * CFD * Thermal plasma reactor * Biomass * Gasification * Syngas Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.355, year: 2016 https://link.springer.com/article/10.1007/s11090-017-9812-z
An analytical method for neutron thermalization calculations in heterogenous reactors
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1965-01-01
It is well known that the use of the diffusion approximation for studying neutron thermalization in heterogeneous reactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations
Thermal hydraulic and safety analyses for Pakistan Research Reactor-1
International Nuclear Information System (INIS)
Bokhari, I.H.; Israr, M.; Pervez, S.
1999-01-01
Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)
Revision of construction plan for advanced thermal demonstration reactor
International Nuclear Information System (INIS)
1996-01-01
The Federation of Electric Power Companies demanded the revision of the construction plan for the advanced thermal demonstration reactor, which is included in the 'Long term plan on the research, development and utilization of atomic energy' decided by the Atomic Energy Commission in 1994, for economical reason. The Atomic Energy Commission carried out the deliberation on this demand. It was found that the cost of construction increases to 580 billion yen, and the cost of electric power generation increases three times as high as that of LWRs. The role as the reactor that utilizes MOX fuel can be substituted by LWRs. The relation of trust with the local town must be considered. In view of these circumstances, it is judged that the stoppage of the construction plan is appropriate. It is necessary to investigate the substitute plan for the stoppage, and the viewpoints of investigating the substitute plan, the examination of the advanced BWR with all MOX fuel core and the method of advancing its construction are considered. On the research and development related to advanced thermal reactors, the research and development contributing to the advance of nuclear fuel recycling are advanced, and the prototype reactor 'Fugen' is utilized. (K.I.)
Thermal barrier and support for nuclear reactor fuel core
International Nuclear Information System (INIS)
Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.
1987-01-01
A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads
Two-dimensional confinement of heavy fermions
International Nuclear Information System (INIS)
Shishido, Hiroaki; Shibauchi, Takasada; Matsuda, Yuji; Terashima, Takahito
2010-01-01
Metallic systems with the strongest electron correlations are realized in certain rare-earth and actinide compounds whose physics are dominated by f-electrons. These materials are known as heavy fermions, so called because the effective mass of the conduction electrons is enhanced via correlation effects up to as much as several hundreds times the free electron mass. To date the electronic structure of all heavy-fermion compounds is essentially three-dimensional. Here we report on the first realization of a two-dimensional heavy-fermion system, where the dimensionality is adjusted in a controllable fashion by fabricating heterostructures using molecular beam epitaxy. The two-dimensional heavy fermion system displays striking deviations from the standard Fermi liquid low-temperature electronic properties. (author)
Two-dimensional sensitivity calculation code: SENSETWO
International Nuclear Information System (INIS)
Yamauchi, Michinori; Nakayama, Mitsuo; Minami, Kazuyoshi; Seki, Yasushi; Iida, Hiromasa.
1979-05-01
A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)
Two-dimensional ranking of Wikipedia articles
Zhirov, A. O.; Zhirov, O. V.; Shepelyansky, D. L.
2010-10-01
The Library of Babel, described by Jorge Luis Borges, stores an enormous amount of information. The Library exists ab aeterno. Wikipedia, a free online encyclopaedia, becomes a modern analogue of such a Library. Information retrieval and ranking of Wikipedia articles become the challenge of modern society. While PageRank highlights very well known nodes with many ingoing links, CheiRank highlights very communicative nodes with many outgoing links. In this way the ranking becomes two-dimensional. Using CheiRank and PageRank we analyze the properties of two-dimensional ranking of all Wikipedia English articles and show that it gives their reliable classification with rich and nontrivial features. Detailed studies are done for countries, universities, personalities, physicists, chess players, Dow-Jones companies and other categories.
Toward two-dimensional search engines
International Nuclear Information System (INIS)
Ermann, L; Shepelyansky, D L; Chepelianskii, A D
2012-01-01
We study the statistical properties of various directed networks using ranking of their nodes based on the dominant vectors of the Google matrix known as PageRank and CheiRank. On average PageRank orders nodes proportionally to a number of ingoing links, while CheiRank orders nodes proportionally to a number of outgoing links. In this way, the ranking of nodes becomes two dimensional which paves the way for the development of two-dimensional search engines of a new type. Statistical properties of information flow on the PageRank–CheiRank plane are analyzed for networks of British, French and Italian universities, Wikipedia, Linux Kernel, gene regulation and other networks. A special emphasis is done for British universities networks using the large database publicly available in the UK. Methods of spam links control are also analyzed. (paper)
Acoustic phonon emission by two dimensional plasmons
International Nuclear Information System (INIS)
Mishonov, T.M.
1990-06-01
Acoustic wave emission of the two dimensional plasmons in a semiconductor or superconductor microstructure is investigated by using the phenomenological deformation potential within the jellium model. The plasmons are excited by the external electromagnetic (e.m.) field. The power conversion coefficient of e.m. energy into acoustic wave energy is also estimated. It is shown, the coherent transformation has a sharp resonance at the plasmon frequency of the two dimensional electron gas (2DEG). The incoherent transformation of the e.m. energy is generated by ohmic dissipation of 2DEG. The method proposed for coherent phonon beam generation can be very effective for high mobility 2DEG and for thin superconducting layers if the plasmon frequency ω is smaller than the superconducting gap 2Δ. (author). 21 refs, 1 fig
Confined catalysis under two-dimensional materials
Li, Haobo; Xiao, Jianping; Fu, Qiang; Bao, Xinhe
2017-01-01
Small spaces in nanoreactors may have big implications in chemistry, because the chemical nature of molecules and reactions within the nanospaces can be changed significantly due to the nanoconfinement effect. Two-dimensional (2D) nanoreactor formed under 2D materials can provide a well-defined model system to explore the confined catalysis. We demonstrate a general tendency for weakened surface adsorption under the confinement of graphene overlayer, illustrating the feasible modulation of su...
Two-Dimensional Extreme Learning Machine
Directory of Open Access Journals (Sweden)
Bo Jia
2015-01-01
(BP networks. However, like many other methods, ELM is originally proposed to handle vector pattern while nonvector patterns in real applications need to be explored, such as image data. We propose the two-dimensional extreme learning machine (2DELM based on the very natural idea to deal with matrix data directly. Unlike original ELM which handles vectors, 2DELM take the matrices as input features without vectorization. Empirical studies on several real image datasets show the efficiency and effectiveness of the algorithm.
Superintegrability on the two dimensional hyperboloid
International Nuclear Information System (INIS)
Akopyan, E.; Pogosyan, G.S.; Kalnins, E.G.; Miller, W. Jr
1998-01-01
This work is devoted to the investigation of the quantum mechanical systems on the two dimensional hyperboloid which admit separation of variables in at least two coordinate systems. Here we consider two potentials introduced in a paper of C.P.Boyer, E.G.Kalnins and P.Winternitz, which haven't been studied yet. An example of an interbasis expansion is given and the structure of the quadratic algebra generated by the integrals of motion is carried out
Two-dimensional Kagome photonic bandgap waveguide
DEFF Research Database (Denmark)
Nielsen, Jens Bo; Søndergaard, Thomas; Libori, Stig E. Barkou
2000-01-01
The transverse-magnetic photonic-bandgap-guidance properties are investigated for a planar two-dimensional (2-D) Kagome waveguide configuration using a full-vectorial plane-wave-expansion method. Single-moded well-localized low-index guided modes are found. The localization of the optical modes...... is investigated with respect to the width of the 2-D Kagome waveguide, and the number of modes existing for specific frequencies and waveguide widths is mapped out....
International Nuclear Information System (INIS)
Lauer, A.; Schwiegk, H.J.; Wu, T.; Cowan, C.L.
1982-03-01
The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analyses from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution
Mechanical exfoliation of two-dimensional materials
Gao, Enlai; Lin, Shao-Zhen; Qin, Zhao; Buehler, Markus J.; Feng, Xi-Qiao; Xu, Zhiping
2018-06-01
Two-dimensional materials such as graphene and transition metal dichalcogenides have been identified and drawn much attention over the last few years for their unique structural and electronic properties. However, their rise begins only after these materials are successfully isolated from their layered assemblies or adhesive substrates into individual monolayers. Mechanical exfoliation and transfer are the most successful techniques to obtain high-quality single- or few-layer nanocrystals from their native multi-layer structures or their substrate for growth, which involves interfacial peeling and intralayer tearing processes that are controlled by material properties, geometry and the kinetics of exfoliation. This procedure is rationalized in this work through theoretical analysis and atomistic simulations. We propose a criterion to assess the feasibility for the exfoliation of two-dimensional sheets from an adhesive substrate without fracturing itself, and explore the effects of material and interface properties, as well as the geometrical, kinetic factors on the peeling behaviors and the torn morphology. This multi-scale approach elucidates the microscopic mechanism of the mechanical processes, offering predictive models and tools for the design of experimental procedures to obtain single- or few-layer two-dimensional materials and structures.
Development of thermal hydraulic evaluation code for CANDU reactors
International Nuclear Information System (INIS)
Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun
2004-02-01
To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted
International Nuclear Information System (INIS)
Iranshahi, Davood; Saeedi, Reza; Azizi, Kolsoom; Nategh, Mahshid
2017-01-01
Highlights: • A novel thermally coupled reactor in CCR naphtha reforming process is modeled. • The required heat of Naphtha process is attained with toluene hydrodealkylation. • A new kinetic model involving 32 pseudo-component and 84 reactions is proposed. • The aromatics and hydrogen production increase 19% and 23%, respectively. - Abstract: Due to the importance of catalytic naphtha reforming process in refineries, development of this process to attain the highest yield of desired products is crucial. In this study, continuous catalyst regeneration naphtha reforming process with radial flow is coupled with hydrodealkylation of toluene to prevent energy loss while enhancing aromatics and hydrogen yields. In this coupled process, heat is transferred between hot and cold sections (from hydrodealkylation of toluene to catalytic naphtha reforming process) using the process integration method. A steady-state two-dimensional model, which considers coke formation on the catalyst pellets, is developed and 32 pseudo-components with 84 reactions are investigated. Kinetic model utilized for HDA process is homogeneous and non-catalytic. The modeling results reveal an approximate increase of 19% and 23% in aromatics and hydrogen molar flow rates, respectively, in comparison with conventional naphtha reforming process. The improvement in aromatics production evidently indicates that HDA is a suitable process to be coupled with naphtha reforming.
Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors
International Nuclear Information System (INIS)
Baek, Won Pil; Song, C. H.; Kim, Y. S.
2007-02-01
The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted
Positron annihilation studies on reactor irradiated and thermal annealed ferrocene
International Nuclear Information System (INIS)
Marques Netto, A.; Carvalho, R.S.; Magalhaes, W.F.; Sinisterra, R.D.
1996-01-01
Retention and thermal annealing following (n, γ) reaction in solid ferrocene, Fe(C 5 H 5 ) 2 , were studied by positron annihilation lifetime spectroscopy (PAL). Positronium (Ps) formation was observed in the non-irradiated compound with a probability or intensity (I 3 ) of 30%. Upon irradiation of the compound with thermal neutrons in a nuclear reactor, I 3 decreases with increasing irradiation time. Thermal treatment again increases I 3 values from 16% to 25%, revealing an important proportion of molecular reformation without variation of the ortho-positronium lifetime (τ 3 ). These results point out the major influence of the electronic structure as determining the Ps yields in the pure complex. In the irradiated and non irradiated complexes the results are satisfactorily explained on the basis of the spur model. (orig.)
International Nuclear Information System (INIS)
1980-01-01
Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base
C.E.C. - cod for calculus of the evolution fuel for thermal reactors
International Nuclear Information System (INIS)
Biciolla, L.; Marcu, G.; Mociornita, G.
1975-01-01
The study of ''burnup'' into thermal reactor involves two main aspects: the economic one and another regarding the reactor operation, its stability and control. In the CEC-code written in FORTRAN IV language was analysed the change of the isotopic composition of nuclear fuel from thermal reactor during its operation
Methods and tools to detect thermal noise in fast reactors
International Nuclear Information System (INIS)
Motta, M.; Giovannini, R.
1985-07-01
The Specialists' Meeting on ''Methods and Tools to Detect Thermal Noise in Fast Reactors'' was held in Bologna on 8-10 October 1984. The meeting was hosted by the ENEA and was sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors. 17 participants attended the meeting from France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, Joint Research Centre of CEC and from IAEA. The meeting was presided over by Prof. Mario Motta of Italy. The purpose of the meeting was to review and discuss methods and tools for temperature noise detection and related analysis as a potential means for detecting local blockages in fuel and blanket subassemblies and other faults in LMFBR. The meeting was divided into four technical sessions as follows: 1. National review presentations on application purposes and research activities for thermal noise detection. (5 papers); 2. Detection instruments and electronic equipment for temperature measurements in fast reactors. (5 papers); 3. Physical models. (2 papers); 4. Signal processing techniques. (3 papers). A separate abstract was prepared for each of these papers
Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril
2017-07-01
There are two type of High Temperature Gas Reactor (HTGR), prismatic and pebble bed. Pebble Bed type has unique configuration because the fuels are randomly distributed inside the reactor core. In term of safety features, Pebble Bed Reactor (PBR) is one of the most promising reactor type in avoiding severe nuclear accidents. In order to analyze heat transfer and safety of this reactor type, a computer code is now under development. As a first step, calculation method proposed by Stroh [1] is adopted. An approach has been made to treat randomly distributed pebble balls contains fissile material inside the reactor core as a porous medium. Helium gas act as coolant on the reactor system are carrying heat flowing in the area between the pebble balls. Several parameters and constants are taken into account in the new developed code. Progress of the development of the code especially comparison of several thermal conductivity constants for a certain PBR-case are reported in the present study.
International Nuclear Information System (INIS)
Vladimir Ya Kumaev; Andrei A Lebezov; Victor V Alexeev
2005-01-01
Full text of publication follows: The report is devoted to the development and application of the two-dimensional MASKA-LM computer code intended for numerical calculations of lead coolant flows, temperatures and transport of impurities in BREST-type reactors of the integral design. The description of heat and mass transfer in liquid metal systems, proceeding in the coolant and at the interface 'coolant - structural materials', is a complex problem involving the joint simulation of thermal-hydraulic, physical and chemical processes in view of the real configuration of the reactor circuit. The report presents the state-of-the-art in the development of the two-dimensional code MASKA-LM and the results of trial calculations of heat and mass transfer in the primary circuit of the lead cooled reactor. The set of governing equations to be solved is based on the porous body model and describes the thermal-hydraulic processes in the reactor as a whole. The numerical method for solution of the governing equations is discussed. To check the code workability and study the technique by the way of solution of a particular task, calculations were performed in reference to the chosen version of the lead cooled BREST reactor under design. The examined domain of the reactor was simulated by a porous body with the parameters corresponding to those of the real reactor medium in terms of heat generation, resistance and the geometry of the hydraulic path of coolant. Analysis of the calculated two-dimensional fields of velocities, pressure and temperatures shows the existence of a complex coolant flow with stagnant and vortex zones. A nonuniform distribution of the coolant flow rate along the core radius was obtained. The results of calculations of the impurity transport of iron, oxygen and magnetite in the primary reactor circuit are discussed as well. The developed code MASKA-LM allows one to evaluate the issue of components of structural materials into coolant as impurities, their
Development potential for thermal reactors and their fuel cycles
International Nuclear Information System (INIS)
Rogers, J.T.; Dodds, H.L. Jr.; Florido, P.C.; Gat, U.; Kondo, S.; Spinks, N.S.
1997-01-01
Water-cooled reactors represent the only types which have reached widespread commercial use up to the present day. Given the plentiful supply of uranium in the world today, this situation might be expected to continue for some time into the future. Nevertheless, for different reasons several countries consider that either new reactor types should be developed or that existing types should be improved substantially. The predominant reason in the short term is to improve the competitive position of nuclear energy supply versus fossil energy. In the longer term, regional and national fuel supply independence may become the dominant driving forces. This paper outlines several possible means for responding to these driving forces. It is not meant to include an exhaustive list of all possibilities, but only to illustrate some alternative routes. These routes range from enhancement of existing reactor concepts to combination of nuclear with fossil systems, and finally to the introduction of radically new thermal reactor concepts. Each of these has its obvious advantages and disadvantages and will come forward or will recede depending on technical feasibility, economics, long-term sustainability, and national policy. (author)
Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling
Energy Technology Data Exchange (ETDEWEB)
Travis, Adam R [ORNL
2014-05-01
A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.
Nuclear data for the calculation of thermal reactor reactivity coefficients
International Nuclear Information System (INIS)
1989-01-01
On its 15th meeting in Vienna, 16-20 June 1986, the International Nuclear Data Committee (INDC) considered it important to review the accuracy with which changes in thermal reactor reactivity resulting from changes in temperature and coolant density can be predicted. It was noted that reactor physicists in several countries had to adjust the thermal neutron cross-section data base in order to reproduce measured reactivity coefficients. Consequently, it appeared to be essential to examine the consistency of the integral and differential cross-section data and to make all the information available which has a bearing on reactivity coefficient prediction. Following the recommendation of the INDC, the Nuclear Data Section of the International Atomic Energy Agency, therefore, convened the Advisory Group Meeting on Nuclear Data for the Calculation of Thermal Reaction Reactivity Coefficients, in Vienna, Austria, 7-10 Dec. 1987. The Conclusions and Recommendations of the meeting together with the papers presented, are submitted in the present document. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs
Thermal hydraulic and neutronic interaction in the rotating bed reactor
International Nuclear Information System (INIS)
Lee, C.C.
1986-01-01
Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors
Vector (two-dimensional) magnetic phenomena
International Nuclear Information System (INIS)
Enokizono, Masato
2002-01-01
In this paper, some interesting phenomena were described from the viewpoint of two-dimensional magnetic property, which is reworded with the vector magnetic property. It shows imperfection of conventional magnetic property and some interested phenomena were discovered, too. We found magnetic materials had the strong nonlinearity both magnitude and spatial phase due to the relationship between the magnetic field strength H-vector and the magnetic flux density B-vector. Therefore, magnetic properties should be defined as the vector relationship. Furthermore, the new Barukhausen signal was observed under rotating flux. (Author)
Two-dimensional Semiconductor-Superconductor Hybrids
DEFF Research Database (Denmark)
Suominen, Henri Juhani
This thesis investigates hybrid two-dimensional semiconductor-superconductor (Sm-S) devices and presents a new material platform exhibiting intimate Sm-S coupling straight out of the box. Starting with the conventional approach, we investigate coupling superconductors to buried quantum well....... To overcome these issues we integrate the superconductor directly into the semiconducting material growth stack, depositing it in-situ in a molecular beam epitaxy system under high vacuum. We present a number of experiments on these hybrid heterostructures, demonstrating near unity interface transparency...
Binding energy of two-dimensional biexcitons
DEFF Research Database (Denmark)
Singh, Jai; Birkedal, Dan; Vadim, Lyssenko
1996-01-01
Using a model structure for a two-dimensional (2D) biexciton confined in a quantum well, it is shown that the form of the Hamiltonian of the 2D biexciton reduces into that of an exciton. The binding energies and Bohr radii of a 2D biexciton in its various internal energy states are derived...... analytically using the fractional dimension approach. The ratio of the binding energy of a 2D biexciton to that of a 2D exciton is found to be 0.228, which agrees very well with the recent experimental value. The results of our approach are compared with those of earlier theories....
Airy beams on two dimensional materials
Imran, Muhammad; Li, Rujiang; Jiang, Yuyu; Lin, Xiao; Zheng, Bin; Dehdashti, Shahram; Xu, Zhiwei; Wang, Huaping
2018-05-01
We propose that quasi-transverse-magnetic (quasi-TM) Airy beams can be supported on two dimensional (2D) materials. By taking graphene as a typical example, the solution of quasi-TM Airy beams is studied under the paraxial approximation. The analytical field intensity in a bilayer graphene-based planar plasmonic waveguide is confirmed by the simulation results. Due to the tunability of the chemical potential of graphene, the self-accelerating behavior of the quasi-TM Airy beam can be steered effectively. 2D materials thus provide a good platform to investigate the propagation of Airy beams.
Two-dimensional heat flow apparatus
McDougall, Patrick; Ayars, Eric
2014-06-01
We have created an apparatus to quantitatively measure two-dimensional heat flow in a metal plate using a grid of temperature sensors read by a microcontroller. Real-time temperature data are collected from the microcontroller by a computer for comparison with a computational model of the heat equation. The microcontroller-based sensor array allows previously unavailable levels of precision at very low cost, and the combination of measurement and modeling makes for an excellent apparatus for the advanced undergraduate laboratory course.
Approaches to passive safety in advanced thermal reactors
International Nuclear Information System (INIS)
Moses, D.L.
1986-01-01
Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time
Magnet system for a thermal barrier Tandem Mirror Reactor
International Nuclear Information System (INIS)
Kim, N.S.; Conn, R.W.
1981-01-01
The magnet system for a thermal barrier D-D tandem mirror reactor has been studied as part of the UCLA tandem mirror reactor design study SATYR. Three main considerations in designing the SATYR magnet system are to obtain the desired field strength variation throughout the system, to have proper space for plasma and neutron shielding, and to satisfy the MHD stability to achieve maximum central cell /beta/. Due to the importance and the complexity, the 'internal' field reversal magnet is the main concern in the entire magnet system for SATYR. Two different magnet designs, a non-uniform current density solenoid and a higher-order solenoid, are discussed. Coil levitation for the internal field reversal magnet has been analyzed
The plutonium utilization in thermal and fast reactor in Japan
International Nuclear Information System (INIS)
Amanuma, T.; Uematsu, K.
1977-01-01
The nuclear power development in Japan is rather extensive one, and the installed nuclear power capacity is expected to reach 49,000 MWe by 1985 and possibly to reach 170,000 MWe by 2000 according to a prediction. Currently istalled nuclear power is mainly based on Light Water type Reactor, and this trend is expected to persist for the time-being. The plutonium produced by LWR will be accumulated to 20 tons by 1985 and to more than 200 tons by 2000. If the produced plutonium will simply be stored, it will raise the economic pressure to utilities and the management and physical protection problems associated with plutonium storing. Therefore, it is not too wise simply to store plutonium in a locked vault. In Japan, there are three ways of solving these problems which are currently worked out. There is no doubt that the best solution is to use plutonium in fast reactors. To reach this goal, an Experimental Fast Reactor ''JOYO''has been constructed and it is waiting for criticality in very near future. A prototype fast breeder reactor ''MONJU'', which is designed for about 300 MWe, is nearing to the last stage of the design work. The start of its construction will take place in a few yesars. The domonstration fast breeder reactor will come next to ''MONJU'' and the large scale commercial use of fast breeder reactor is expected to start around 1995. To anwer the near-term need for plutonium utilization, two technologies, which are equally important to Japan, are currently developed. One is the recycle use of plutonium into LWR. This technology has long been jointly developed by research organizations and utilities. Some of fuel irradiation data are already obtained and the physics study has also been extensive. The application of this technology is expected to start about 1987. The other is to burn plutonium in an Advanced Thermal Reactor (D 2 O moderated, Boiling Water Cooled) which shows better characteristics of using plutonium. The 160 MWe ''Fugen'' is a prototype
Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics
International Nuclear Information System (INIS)
Santos Bastos, W. dos
1995-01-01
These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods
Transmutation of Thermocouples in Thermal and Fast Nuclear Reactors
International Nuclear Information System (INIS)
Scervini, M.; Rae, C.; Lindley, B.
2013-06-01
Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. Their role is fundamental for the control of current nuclear reactors and for the development of the nuclear technology needed for the implementation of GEN IV nuclear reactors. When used for in-core measurements thermocouples are strongly affected not only by high temperatures, but also by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition in the thermoelements and, as a consequence, a time dependent drift in the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. In this work, undertaken as part of the European project METROFISSION, the change in composition occurring in irradiated thermocouples has been calculated using the software ORIGEN 2.2. Several thermocouples have been considered, including Nickel based thermocouples (type K and type N), Tungsten based thermocouples (W-5%Re vs W-26%Re and W- 3%Re vs W-25%Re), Platinum based thermocouples (type S and Platinum vs Palladium) and Molybdenum vs Niobium thermocouples. The transmutation induced by both thermal flux and fast flux has been calculated. Thermocouples undergo more pronounced transmutation in thermal fluxes rather than in fast fluxes, as the neutron cross section of an element is higher for thermal energies. Nickel based thermocouples have a minimal change in composition, while Platinum based and Tungsten based thermocouples experience a very significant transmutation. The use of coatings deposited on the sheath of a thermocouple has been considered as a mean to reduce the neutron flux the thermoelements inside the thermocouple sheath
Thermal reactor benchmark testing of 69 group library
International Nuclear Information System (INIS)
Liu Guisheng; Wang Yaoqing; Liu Ping; Zhang Baocheng
1994-01-01
Using a code system NSLINK, AMPX master library in WIMS 69 groups structure are made from nuclides relating to 4 newest evaluated nuclear data libraries. Some integrals of 10 thermal reactor benchmark assemblies recommended by the U.S. CSEWG are calculated using rectified PASC-1 code system and compared with foreign results, the authors results are in good agreement with others. 69 group libraries of evaluated data bases in TPFAP interface file are generated with NJOY code system. The k ∞ values of 6 cell lattice assemblies are calculated by the code CBM. The calculated results are analysed and compared
In-reactor creep of zirconium alloys by thermal spikes
International Nuclear Information System (INIS)
Ibrahim, E.F.
1975-01-01
The size and duration of thermal spikes from fast neutrons have been calculated for zirconium alloys, showing that spikes up to 1.8 nm radius may exist for 2 x 10 -11 s at greater than melting point, at 570K ambient temperature. Creep rates have been calculated assuming that the elastic strain from the applied stress relaxes in the volume of the spikes (by preferential loop alignment or modification of an existing dislocation network). The calculated rates are consistent with strain rates observed in long term tests-in-reactor, if spike lifetimes are 2 to 2.5 x 10 -11 s. (Auth.)
Thermal-hydraulic considerations for particle bed reactors
Benenati, R.; Araj, K. J.; Horn, F.
In the design of particle bed reactor (PBR) cores, consideration must be given to the gas coolant channels and their configuration. Neutronics analysis provides the relative volume fractions of the component materials, but these must be arranged in such a manner as to allow proper cooling of all components by the gas flow at relatively low pressure drops. The thermal hydraulic aspects of this problem are addressed. A description of the computer model used in the analysis of the steady state condition is also included. Blowdown tests on hot particle bed fuel elements were carried out and are described.
Pressure thermal shock analysis for nuclear reactor pressure vessel
International Nuclear Information System (INIS)
Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.
2015-01-01
The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)
Decoherence in two-dimensional quantum walks
International Nuclear Information System (INIS)
Oliveira, A. C.; Portugal, R.; Donangelo, R.
2006-01-01
We analyze the decoherence in quantum walks in two-dimensional lattices generated by broken-link-type noise. In this type of decoherence, the links of the lattice are randomly broken with some given constant probability. We obtain the evolution equation for a quantum walker moving on two-dimensional (2D) lattices subject to this noise, and we point out how to generalize for lattices in more dimensions. In the nonsymmetric case, when the probability of breaking links in one direction is different from the probability in the perpendicular direction, we have obtained a nontrivial result. If one fixes the link-breaking probability in one direction, and gradually increases the probability in the other direction from 0 to 1, the decoherence initially increases until it reaches a maximum value, and then it decreases. This means that, in some cases, one can increase the noise level and still obtain more coherence. Physically, this can be explained as a transition from a decoherent 2D walk to a coherent 1D walk
Two-Dimensional Theory of Scientific Representation
Directory of Open Access Journals (Sweden)
A Yaghmaie
2013-03-01
Full Text Available Scientific representation is an interesting topic for philosophers of science, many of whom have recently explored it from different points of view. There are currently two competing approaches to the issue: cognitive and non-cognitive, and each of them claims its own merits over the other. This article tries to provide a hybrid theory of scientific representation, called Two-Dimensional Theory of Scientific Representation, which has the merits of the two accounts and is free of their shortcomings. To do this, we will argue that although scientific representation needs to use the notion of intentionality, such a notion is defined and realized in a simply structural form contrary to what cognitive approach says about intentionality. After a short introduction, the second part of the paper is devoted to introducing theories of scientific representation briefly. In the third part, the structural accounts of representation will be criticized. The next step is to introduce the two-dimensional theory which involves two key components: fixing and structural fitness. It will be argued that fitness is an objective and non-intentional relation, while fixing is intentional.
Strain-engineered growth of two-dimensional materials.
Ahn, Geun Ho; Amani, Matin; Rasool, Haider; Lien, Der-Hsien; Mastandrea, James P; Ager Iii, Joel W; Dubey, Madan; Chrzan, Daryl C; Minor, Andrew M; Javey, Ali
2017-09-20
The application of strain to semiconductors allows for controlled modification of their band structure. This principle is employed for the manufacturing of devices ranging from high-performance transistors to solid-state lasers. Traditionally, strain is typically achieved via growth on lattice-mismatched substrates. For two-dimensional (2D) semiconductors, this is not feasible as they typically do not interact epitaxially with the substrate. Here, we demonstrate controlled strain engineering of 2D semiconductors during synthesis by utilizing the thermal coefficient of expansion mismatch between the substrate and semiconductor. Using WSe 2 as a model system, we demonstrate stable built-in strains ranging from 1% tensile to 0.2% compressive on substrates with different thermal coefficient of expansion. Consequently, we observe a dramatic modulation of the band structure, manifested by a strain-driven indirect-to-direct bandgap transition and brightening of the dark exciton in bilayer and monolayer WSe 2 , respectively. The growth method developed here should enable flexibility in design of more sophisticated devices based on 2D materials.Strain engineering is an essential tool for modifying local electronic properties in silicon-based electronics. Here, Ahn et al. demonstrate control of biaxial strain in two-dimensional materials based on the growth substrate, enabling more complex low-dimensional electronics.
Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model
Directory of Open Access Journals (Sweden)
Reza Akbari
2017-08-01
Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.
A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor
Energy Technology Data Exchange (ETDEWEB)
Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)
1998-03-01
The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.
Need for nuclear data for thermal neutron reactors
International Nuclear Information System (INIS)
Bouchard, J.; Golinelli, C.; Tellier, H.
1983-01-01
The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)
Two-Dimensional Tellurene as Excellent Thermoelectric Material
Sharma, Sitansh
2018-04-20
We study the thermoelectric properties of two-dimensional tellurene by first-principles calculations and semiclassical Boltzmann transport theory. The HSE06 hybrid functional results in a moderate direct band gap of 1.48 eV at the Γ point. A high room temperature Seebeck coefficient (Sxx = 0.38 mV/K, Syy = 0.36 mV/K) is combined with anisotropic lattice thermal conductivity (κxxl = 0.43 W/m K, κyyl = 1.29 W/m K). Phonon band structures demonstrate a key role of optical phonons in the record low thermal conductivity that leads to excellent thermoelectric performance of tellurene. At room temperature and moderate hole doping of 1.2 × 10–11 cm–2, for example, a figure of merit of ZTxx = 0.8 is achieved.
Transient thermal characteristics of a core channel in a molten salt reactor
International Nuclear Information System (INIS)
Sakashita, H.; Ishiguro, R.; Sugiyama, K.
1987-01-01
The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)
International Nuclear Information System (INIS)
Penndorf, K.
1976-04-01
Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time
In-place thermal annealing of nuclear reactor pressure vessels
International Nuclear Information System (INIS)
Server, W.L.
1985-04-01
Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)
Thermal energy and bootstrap current in fusion reactor plasmas
International Nuclear Information System (INIS)
Becker, G.
1993-01-01
For DT fusion reactors with prescribed alpha particle heating power P α , plasma volume V and burn temperature i > ∼ 10 keV specific relations for the thermal energy content, bootstrap current, central plasma pressure and other quantities are derived. It is shown that imposing P α and V makes these relations independent of the magnitudes of the density and temperature, i.e. they only depend on P α , V and shape factors or profile parameters. For model density and temperature profiles analytic expressions for these shape factors and for the factor C bs in the bootstrap current formula I bs ∼ C bs (a/R) 1/2 β p I p are given. In the design of next-step devices and fusion reactors, the fusion power is a fixed quantity. Prescription of the alpha particle heating power and plasma volume results in specific relations which can be helpful for interpreting computer simulations and for the design of fusion reactors. (author) 5 refs
Thermal-hydraulic modeling of porous bed reactors
International Nuclear Information System (INIS)
Araj, K.J.; Nourbakhsh, H.P.
1987-01-01
Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures
Isotopes accumulation in the thermal column of TRIGA reactor
International Nuclear Information System (INIS)
Iorgulis, C.; Diaconu, D.; Gugiu, D.; Csaba, R.
2013-01-01
The correlation of impurity observed in the virgin graphite and radionuclide content and activities measured in the irradiated graphite needs to know the irradiated history. This is a challenging process if impurity content and irradiation conditions are not accurately known. This is the case of the irradiated graphite in the thermal column of Institute for Nuclear Research Pitesti (INR)14 MW TRIGA reactor. To overcome incomplete impurity content and the unknown position in the column of the measured irradiated graphite available for characterisation and comparison, a set of preliminary simulations were performed. Following Eu 152 /Eu 154 ration they allowed the estimation of an impurity content and irradiation conditions leading to measured activities. Based on these data the radio-isotope accumulation in different positions in the thermal column was predicted. Modelling performed by INR used advanced prediction packages (e.g. WIMS, MCNP ORIGEN-S from Scale 5) to assess the isotopic content of MTR graphite types with irradiation history specific for a TRIGA research reactor. Some certain calculations points from the column were selected in order to model the burnup and isotopes productions using ORIGEN from SCALE code system. (authors)
International Nuclear Information System (INIS)
Nomura, Yasushi; Someya, Hiroyuki; Ito, Haruhiko.
1992-11-01
Capsules for irradiation tests in the JMTR (Japan Materials Testing Reactor), consist of irradiation specimens surrounded by a cladding tube, holders, an inner tube and a container tube (from 30mm to 65mm in diameter). And the annular gaps between these structural materials in the capsule are filled with liquids or gases. Cooling of the capsule is done by reactor primary coolant flowing down outside the capsule. Most of the heat generated by fission in fuel specimens and gamma absorption in structural materials is directed radially to the capsule container outer surface. In thermal performance calculations for capsule design, an one(r)-dimensional heat transfer computer code entitled (Generalyzed Gap Temperature Calculation), GENGTC, originally developed in Oak Ridge National Laboratory, U.S.A., has been frequently used. In designing a capsule, are needed many cases of parametric calculations with respect to changes materials and gap sizes. And in some cases, two(r,z)-dimensional heat transfer calculations are needed for irradiation test capsules with short length fuel rods. Recently the authors improved the original one-dimensional code GENGTC, (1) to simplify preparation of input data, (2) to perform automatic calculations for parametric survey based on design temperatures, ect. Moreover, the computer code has been improved to perform r-z two-dimensional heat transfer calculation. This report describes contents of the preparation of the one-dimensional code GENGTC and the improvement for the two-dimensional code GENGTC-2, together with their code manuals. (author)
Two-dimensional simulation of sintering process
International Nuclear Information System (INIS)
Vasconcelos, Vanderley de; Pinto, Lucio Carlos Martins; Vasconcelos, Wander L.
1996-01-01
The results of two-dimensional simulations are directly applied to systems in which one of the dimensions is much smaller than the others, and to sections of three dimensional models. Moreover, these simulations are the first step of the analysis of more complex three-dimensional systems. In this work, two basic features of the sintering process are studied: the types of particle size distributions related to the powder production processes and the evolution of geometric parameters of the resultant microstructures during the solid-state sintering. Random packing of equal spheres is considered in the sintering simulation. The packing algorithm does not take into account the interactive forces between the particles. The used sintering algorithm causes the densification of the particle set. (author)
Two dimensional generalizations of the Newcomb equation
International Nuclear Information System (INIS)
Dewar, R.L.; Pletzer, A.
1989-11-01
The Bineau reduction to scalar form of the equation governing ideal, zero frequency linearized displacements from a hydromagnetic equilibrium possessing a continuous symmetry is performed in 'universal coordinates', applicable to both the toroidal and helical cases. The resulting generalized Newcomb equation (GNE) has in general a more complicated form than the corresponding one dimensional equation obtained by Newcomb in the case of circular cylindrical symmetry, but in this cylindrical case , the equation can be transformed to that of Newcomb. In the two dimensional case there is a transformation which leaves the form of the GNE invariant and simplifies the Frobenius expansion about a rational surface, especially in the limit of zero pressure gradient. The Frobenius expansions about a mode rational surface is developed and the connection with Hamiltonian transformation theory is shown. 17 refs
Pressure of two-dimensional Yukawa liquids
International Nuclear Information System (INIS)
Feng, Yan; Wang, Lei; Tian, Wen-de; Goree, J; Liu, Bin
2016-01-01
A simple analytic expression for the pressure of a two-dimensional Yukawa liquid is found by fitting results from a molecular dynamics simulation. The results verify that the pressure can be written as the sum of a potential term which is a simple multiple of the Coulomb potential energy at a distance of the Wigner–Seitz radius, and a kinetic term which is a multiple of the one for an ideal gas. Dimensionless coefficients for each of these terms are found empirically, by fitting. The resulting analytic expression, with its empirically determined coefficients, is plotted as isochores, or curves of constant area. These results should be applicable to monolayer dusty plasmas. (paper)
Two dimensional nanomaterials for flexible supercapacitors.
Peng, Xu; Peng, Lele; Wu, Changzheng; Xie, Yi
2014-05-21
Flexible supercapacitors, as one of most promising emerging energy storage devices, are of great interest owing to their high power density with great mechanical compliance, making them very suitable as power back-ups for future stretchable electronics. Two-dimensional (2D) nanomaterials, including the quasi-2D graphene and inorganic graphene-like materials (IGMs), have been greatly explored to providing huge potential for the development of flexible supercapacitors with higher electrochemical performance. This review article is devoted to recent progresses in engineering 2D nanomaterials for flexible supercapacitors, which survey the evolution of electrode materials, recent developments in 2D nanomaterials and their hybrid nanostructures with regulated electrical properties, and the new planar configurations of flexible supercapacitors. Furthermore, a brief discussion on future directions, challenges and opportunities in this fascinating area is also provided.
Geometrical aspects of solvable two dimensional models
International Nuclear Information System (INIS)
Tanaka, K.
1989-01-01
It was noted that there is a connection between the non-linear two-dimensional (2D) models and the scalar curvature r, i.e., when r = -2 the equations of motion of the Liouville and sine-Gordon models were obtained. Further, solutions of various classical nonlinear 2D models can be obtained from the condition that the appropriate curvature two form Ω = 0, which suggests that these models are closely related. This relation is explored further in the classical version by obtaining the equations of motion from the evolution equations, the infinite number of conserved quantities, and the common central charge. The Poisson brackets of the solvable 2D models are specified by the Virasoro algebra. 21 refs
Two-dimensional materials for ultrafast lasers
International Nuclear Information System (INIS)
Wang Fengqiu
2017-01-01
As the fundamental optical properties and novel photophysics of graphene and related two-dimensional (2D) crystals are being extensively investigated and revealed, a range of potential applications in optical and optoelectronic devices have been proposed and demonstrated. Of the many possibilities, the use of 2D materials as broadband, cost-effective and versatile ultrafast optical switches (or saturable absorbers) for short-pulsed lasers constitutes a rapidly developing field with not only a good number of publications, but also a promising prospect for commercial exploitation. This review primarily focuses on the recent development of pulsed lasers based on several representative 2D materials. The comparative advantages of these materials are discussed, and challenges to practical exploitation, which represent good future directions of research, are laid out. (paper)
Two-dimensional phase fraction charts
International Nuclear Information System (INIS)
Morral, J.E.
1984-01-01
A phase fraction chart is a graphical representation of the amount of each phase present in a system as a function of temperature, composition or other variable. Examples are phase fraction versus temperature charts used to characterize specific alloys and as a teaching tool in elementary texts, and Schaeffler diagrams used to predict the amount of ferrite in stainless steel welds. Isothermal-transformation diagrams (TTT diagrams) are examples that give phase (or microconstituent) amount versus temperature and time. The purpose of this communication is to discuss the properties of two-dimensional phase fraction charts in more general terms than have been reported before. It is shown that they can represent multi-component, multiphase equilibria in a way which is easier to read and which contains more information than the isotherms and isopleths of multi-component phase diagrams
Two-dimensional motions of rockets
International Nuclear Information System (INIS)
Kang, Yoonhwan; Bae, Saebyok
2007-01-01
We analyse the two-dimensional motions of the rockets for various types of rocket thrusts, the air friction and the gravitation by using a suitable representation of the rocket equation and the numerical calculation. The slope shapes of the rocket trajectories are discussed for the three types of rocket engines. Unlike the projectile motions, the descending parts of the trajectories tend to be gentler and straighter slopes than the ascending parts for relatively large launching angles due to the non-vanishing thrusts. We discuss the ranges, the maximum altitudes and the engine performances of the rockets. It seems that the exponential fuel exhaustion can be the most potent engine for the longest and highest flights
Two dimensional NMR studies of polysaccharides
International Nuclear Information System (INIS)
Byrd, R.A.; Egan, W.; Summers, M.F.
1987-01-01
Polysaccharides are very important components in the immune response system. Capsular polysaccharides and lipopolysaccharides occupy cell surface sites of bacteria, play key roles in recognition and some have been used to develop vaccines. Consequently, the ability to determine chemical structures of these systems is vital to an understanding of their immunogenic action. The authors have been utilizing recently developed two-dimensional homonuclear and heteronuclear correlation spectroscopy for unambiguous assignment and structure determination of a number of polysaccharides. In particular, the 1 H-detected heteronuclear correlation experiments are essential to the rapid and sensitive determination of these structures. Linkage sites are determined by independent polarization transfer experiments and multiple quantum correlation experiments. These methods permit the complete structure determination on very small amounts of the polysaccharides. They present the results of a number of structural determinations and discuss the limits of these experiments in terms of their applications to polysaccharides
Two-Dimensional Homogeneous Fermi Gases
Hueck, Klaus; Luick, Niclas; Sobirey, Lennart; Siegl, Jonas; Lompe, Thomas; Moritz, Henning
2018-02-01
We report on the experimental realization of homogeneous two-dimensional (2D) Fermi gases trapped in a box potential. In contrast to harmonically trapped gases, these homogeneous 2D systems are ideally suited to probe local as well as nonlocal properties of strongly interacting many-body systems. As a first benchmark experiment, we use a local probe to measure the density of a noninteracting 2D Fermi gas as a function of the chemical potential and find excellent agreement with the corresponding equation of state. We then perform matter wave focusing to extract the momentum distribution of the system and directly observe Pauli blocking in a near unity occupation of momentum states. Finally, we measure the momentum distribution of an interacting homogeneous 2D gas in the crossover between attractively interacting fermions and bosonic dimers.
Two-dimensional electroacoustic waves in silicene
Zhukov, Alexander V.; Bouffanais, Roland; Konobeeva, Natalia N.; Belonenko, Mikhail B.
2018-01-01
In this letter, we investigate the propagation of two-dimensional electromagnetic waves in a piezoelectric medium built upon silicene. Ultrashort optical pulses of Gaussian form are considered to probe this medium. On the basis of Maxwell's equations supplemented with the wave equation for the medium's displacement vector, we obtain the effective governing equation for the vector potential associated with the electromagnetic field, as well as the component of the displacement vector. The dependence of the pulse shape on the bandgap in silicene and the piezoelectric coefficient of the medium was analyzed, thereby revealing a nontrivial triadic interplay between the characteristics of the pulse dynamics, the electronic properties of silicene, and the electrically induced mechanical vibrations of the medium. In particular, we uncovered the possibility for an amplification of the pulse amplitude through the tuning of the piezoelectric coefficient. This property could potentially offer promising prospects for the development of amplification devices for the optoelectronics industry.
Versatile two-dimensional transition metal dichalcogenides
DEFF Research Database (Denmark)
Canulescu, Stela; Affannoukoué, Kévin; Döbeli, Max
), a strategy for the fabrication of 2D heterostructures must be developed. Here we demonstrate a novel approach for the bottom-up synthesis of TMDC monolayers, namely Pulsed Laser Deposition (PLD) combined with a sulfur evaporation beam. PLD relies on the use of a pulsed laser (ns pulse duration) to induce...... material transfer from a solid source (such as a sintered target of MoS2) to a substrate (such as Si or sapphire). The deposition rate in PLD is typically much less than a monolayer per pulse, meaning that the number of MLs can be controlled by a careful selection of the number of laser pulses......Two-dimensional transition metal dichalcogenides (2D-TMDCs), such as MoS2, have emerged as a new class of semiconducting materials with distinct optical and electrical properties. The availability of 2D-TMDCs with distinct band gaps allows for unlimited combinations of TMDC monolayers (MLs...
Two-dimensional heterostructures for energy storage
Energy Technology Data Exchange (ETDEWEB)
Gogotsi, Yury G. [Drexel Univ., Philadelphia, PA (United States); Pomerantseva, Ekaterina [Drexel Univ., Philadelphia, PA (United States)
2017-06-12
Two-dimensional (2D) materials provide slit-shaped ion diffusion channels that enable fast movement of lithium and other ions. However, electronic conductivity, the number of intercalation sites, and stability during extended cycling are also crucial for building high-performance energy storage devices. While individual 2D materials, such as graphene, show some of the required properties, none of them can offer all properties needed to maximize energy density, power density, and cycle life. Here we argue that stacking different 2D materials into heterostructured architectures opens an opportunity to construct electrodes that would combine the advantages of the individual building blocks while eliminating the associated shortcomings. We discuss characteristics of common 2D materials and provide examples of 2D heterostructured electrodes that showed new phenomena leading to superior electrochemical performance. As a result, we also consider electrode fabrication approaches and finally outline future steps to create 2D heterostructured electrodes that could greatly expand current energy storage technologies.
Two-dimensional fourier transform spectrometer
DeFlores, Lauren; Tokmakoff, Andrei
2013-09-03
The present invention relates to a system and methods for acquiring two-dimensional Fourier transform (2D FT) spectra. Overlap of a collinear pulse pair and probe induce a molecular response which is collected by spectral dispersion of the signal modulated probe beam. Simultaneous collection of the molecular response, pulse timing and characteristics permit real time phasing and rapid acquisition of spectra. Full spectra are acquired as a function of pulse pair timings and numerically transformed to achieve the full frequency-frequency spectrum. This method demonstrates the ability to acquire information on molecular dynamics, couplings and structure in a simple apparatus. Multi-dimensional methods can be used for diagnostic and analytical measurements in the biological, biomedical, and chemical fields.
Discrete formulation for two-dimensional multigroup neutron diffusion equations
Energy Technology Data Exchange (ETDEWEB)
Vosoughi, Naser E-mail: vosoughi@mehr.sharif.edu; Salehi, Ali A.; Shahriari, Majid
2003-02-01
The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method.
Discrete formulation for two-dimensional multigroup neutron diffusion equations
International Nuclear Information System (INIS)
Vosoughi, Naser; Salehi, Ali A.; Shahriari, Majid
2003-01-01
The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method
Scaling in nuclear reactor system thermal-hydraulics
International Nuclear Information System (INIS)
D'Auria, F.; Galassi, G.M.
2010-01-01
Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.
Scaling in nuclear reactor system thermal-hydraulics
Energy Technology Data Exchange (ETDEWEB)
D' Auria, F., E-mail: dauria@ing.unipi.i [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Galassi, G.M. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)
2010-10-15
Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.
Equivalency of two-dimensional algebras
International Nuclear Information System (INIS)
Santos, Gildemar Carneiro dos; Pomponet Filho, Balbino Jose S.
2011-01-01
Full text: Let us consider a vector z = xi + yj over the field of real numbers, whose basis (i,j) satisfy a given algebra. Any property of this algebra will be reflected in any function of z, so we can state that the knowledge of the properties of an algebra leads to more general conclusions than the knowledge of the properties of a function. However structural properties of an algebra do not change when this algebra suffers a linear transformation, though the structural constants defining this algebra do change. We say that two algebras are equivalent to each other whenever they are related by a linear transformation. In this case, we have found that some relations between the structural constants are sufficient to recognize whether or not an algebra is equivalent to another. In spite that the basis transform linearly, the structural constants change like a third order tensor, but some combinations of these tensors result in a linear transformation, allowing to write the entries of the transformation matrix as function of the structural constants. Eventually, a systematic way to find the transformation matrix between these equivalent algebras is obtained. In this sense, we have performed the thorough classification of associative commutative two-dimensional algebras, and find that even non-division algebra may be helpful in solving non-linear dynamic systems. The Mandelbrot set was used to have a pictorial view of each algebra, since equivalent algebras result in the same pattern. Presently we have succeeded in classifying some non-associative two-dimensional algebras, a task more difficult than for associative one. (author)
Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-
International Nuclear Information System (INIS)
Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho
1994-07-01
The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)
Prestressed concrete reactor vessel thermal cylinder model study
International Nuclear Information System (INIS)
Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.
1977-01-01
The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a 1 / 6 -scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating
Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems
International Nuclear Information System (INIS)
Song, C. H.; Baek, W. P.; Chung, M. K.
2007-06-01
The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base
Transient thermal creep of nuclear reactor pressure vessel type concretes
International Nuclear Information System (INIS)
Khoury, G.A.
1983-01-01
The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and
Method and apparatus for a combination moving bed thermal treatment reactor and moving bed filter
Energy Technology Data Exchange (ETDEWEB)
Badger, Phillip C.; Dunn, Jr., Kenneth J.
2015-09-01
A moving bed gasification/thermal treatment reactor includes a geometry in which moving bed reactor particles serve as both a moving bed filter and a heat carrier to provide thermal energy for thermal treatment reactions, such that the moving bed filter and the heat carrier are one and the same to remove solid particulates or droplets generated by thermal treatment processes or injected into the moving bed filter from other sources.
Reactor physics analysis of the pin-cell Doppler effect in a thermal nuclear reactor
International Nuclear Information System (INIS)
Kruijf, W.J.M. de.
1995-01-01
This report has also been published as a PhD thesis. It deals with the Doppler effect in thermal nuclear reactors. Especially the behaviour of the reactor in transient conditions is an important issue. During such a transient the radial temperature profile in a fuel pin changes. In this PhD research effective fuel temperatures have been calculated for arbitrary temperature profiles in the fuel pin with the improved slowing-down code ROLAIDS-CPM. A general expression for the effective fuel temperature in a specific fuel pin is found by defining this effective fuel temperature as a weighted sum of the temperatures in different radial fuel zones. Also, the radial power profile in a fuel pin has been calculated by performing detailed burnup calculations, which agree very well with experimental data. (orig.)
Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems
International Nuclear Information System (INIS)
Song, C. H.; Chung, M. K.; Park, C. K. and others
2005-04-01
The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved
Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems
Energy Technology Data Exchange (ETDEWEB)
Song, C. H.; Chung, M. K.; Park, C. K. and others
2005-04-15
The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.
Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)
Energy Technology Data Exchange (ETDEWEB)
Bradley K. Heath
2014-03-01
This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.
Aspects of jamming in two-dimensional athermal frictionless systems.
Reichhardt, C; Reichhardt, C J Olson
2014-05-07
In this work we provide an overview of jamming transitions in two dimensional systems focusing on the limit of frictionless particle interactions in the absence of thermal fluctuations. We first discuss jamming in systems with short range repulsive interactions, where the onset of jamming occurs at a critical packing density and where certain quantities show a divergence indicative of critical behavior. We describe how aspects of the dynamics change as the jamming density is approached and how these dynamics can be explored using externally driven probes. Different particle shapes can produce jamming densities much lower than those observed for disk-shaped particles, and we show how jamming exhibits fragility for some shapes while for other shapes this is absent. Next we describe the effects of long range interactions and jamming behavior in systems such as charged colloids, vortices in type-II superconductors, and dislocations. We consider the effect of adding obstacles to frictionless jamming systems and discuss connections between this type of jamming and systems that exhibit depinning transitions. Finally, we discuss open questions such as whether the jamming transition in all these different systems can be described by the same or a small subset of universal behaviors, as well as future directions for studies of jamming transitions in two dimensional systems, such as jamming in self-driven or active matter systems.
International Nuclear Information System (INIS)
Fernandez-Armas, Sergio; Mesa, Jose L.; Pizarro, Jose L.; Chung, U-Chan; Arriortua, Maria I.; Rojo, Teofilo
2005-01-01
The organically templated (C 6 H 16 N 2 ) 0.5 [M(HPO 3 )F] [M(II)=Fe (1) and Co (2)] compounds have been synthesized by using mild hydrothermal conditions under autogeneous pressure. The crystal structures have been determined from X-ray single-crystal diffraction data. The compounds are isostructural and crystallize in the C2/c monoclinic space group. The unit-cell parameters are a=5.607(1), b=21.276(4), c=11.652(1)A, β=93.74(1) deg. for the iron phase and a=5.5822(7), b=21.325(3), c=11.4910(1)A, β=93.464(9) o for the cobalt compound with Z=4. The crystal structure of these compounds consists of [M(HPO 3 )F] - anionic sheets. The layers are constructed from chains which contain [M 2 O 6 F 3 ] dimeric units linked by fluoride ions. The trans-1,4-diaminocyclohexane cations are placed in the interlayer space. The IR and Raman spectra show the bands corresponding to the phosphite oxoanion and organic dication. The Dq and Racah (B and C) parameters have been calculated from the diffuse reflectance spectra in the visible region. Dq parameter is 790cm -1 for compound (1). For phase (2) the Dq value is 725cm -1 and B and C are 930 and 4100cm -1 , respectively. The thermal evolution of the molar magnetic susceptibilities of these compounds show maxima at 20.0 and 6.0K for the iron(II) and cobalt(II) phases, respectively. These results indicate the existence of antiferromagnetic interactions in both compounds
Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases
Energy Technology Data Exchange (ETDEWEB)
Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1999-12-31
A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)
Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases
Energy Technology Data Exchange (ETDEWEB)
Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)
Electronic Transport in Two-Dimensional Materials
Sangwan, Vinod K.; Hersam, Mark C.
2018-04-01
Two-dimensional (2D) materials have captured the attention of the scientific community due to the wide range of unique properties at nanometer-scale thicknesses. While significant exploratory research in 2D materials has been achieved, the understanding of 2D electronic transport and carrier dynamics remains in a nascent stage. Furthermore, because prior review articles have provided general overviews of 2D materials or specifically focused on charge transport in graphene, here we instead highlight charge transport mechanisms in post-graphene 2D materials, with particular emphasis on transition metal dichalcogenides and black phosphorus. For these systems, we delineate the intricacies of electronic transport, including band structure control with thickness and external fields, valley polarization, scattering mechanisms, electrical contacts, and doping. In addition, electronic interactions between 2D materials are considered in the form of van der Waals heterojunctions and composite films. This review concludes with a perspective on the most promising future directions in this fast-evolving field.
Stress distribution in two-dimensional silos
Blanco-Rodríguez, Rodolfo; Pérez-Ángel, Gabriel
2018-01-01
Simulations of a polydispersed two-dimensional silo were performed using molecular dynamics, with different numbers of grains reaching up to 64 000, verifying numerically the model derived by Janssen and also the main assumption that the walls carry part of the weight due to the static friction between grains with themselves and those with the silo's walls. We vary the friction coefficient, the radii dispersity, the silo width, and the size of grains. We find that the Janssen's model becomes less relevant as the the silo width increases since the behavior of the stresses becomes more hydrostatic. Likewise, we get the normal and tangential stress distribution on the walls evidencing the existence of points of maximum stress. We also obtained the stress matrix with which we observe zones of concentration of load, located always at a height around two thirds of the granular columns. Finally, we observe that the size of the grains affects the distribution of stresses, increasing the weight on the bottom and reducing the normal stress on the walls, as the grains are made smaller (for the same total mass of the granulate), giving again a more hydrostatic and therefore less Janssen-type behavior for the weight of the column.
Asymptotics for Two-dimensional Atoms
DEFF Research Database (Denmark)
Nam, Phan Thanh; Portmann, Fabian; Solovej, Jan Philip
2012-01-01
We prove that the ground state energy of an atom confined to two dimensions with an infinitely heavy nucleus of charge $Z>0$ and $N$ quantum electrons of charge -1 is $E(N,Z)=-{1/2}Z^2\\ln Z+(E^{\\TF}(\\lambda)+{1/2}c^{\\rm H})Z^2+o(Z^2)$ when $Z\\to \\infty$ and $N/Z\\to \\lambda$, where $E^{\\TF}(\\lambd......We prove that the ground state energy of an atom confined to two dimensions with an infinitely heavy nucleus of charge $Z>0$ and $N$ quantum electrons of charge -1 is $E(N,Z)=-{1/2}Z^2\\ln Z+(E^{\\TF}(\\lambda)+{1/2}c^{\\rm H})Z^2+o(Z^2)$ when $Z\\to \\infty$ and $N/Z\\to \\lambda$, where $E......^{\\TF}(\\lambda)$ is given by a Thomas-Fermi type variational problem and $c^{\\rm H}\\approx -2.2339$ is an explicit constant. We also show that the radius of a two-dimensional neutral atom is unbounded when $Z\\to \\infty$, which is contrary to the expected behavior of three-dimensional atoms....
Seismic isolation of two dimensional periodic foundations
International Nuclear Information System (INIS)
Yan, Y.; Mo, Y. L.; Laskar, A.; Cheng, Z.; Shi, Z.; Menq, F.; Tang, Y.
2014-01-01
Phononic crystal is now used to control acoustic waves. When the crystal goes to a larger scale, it is called periodic structure. The band gaps of the periodic structure can be reduced to range from 0.5 Hz to 50 Hz. Therefore, the periodic structure has potential applications in seismic wave reflection. In civil engineering, the periodic structure can be served as the foundation of upper structure. This type of foundation consisting of periodic structure is called periodic foundation. When the frequency of seismic waves falls into the band gaps of the periodic foundation, the seismic wave can be blocked. Field experiments of a scaled two dimensional (2D) periodic foundation with an upper structure were conducted to verify the band gap effects. Test results showed the 2D periodic foundation can effectively reduce the response of the upper structure for excitations with frequencies within the frequency band gaps. When the experimental and the finite element analysis results are compared, they agree well with each other, indicating that 2D periodic foundation is a feasible way of reducing seismic vibrations.
Two-dimensional topological photonic systems
Sun, Xiao-Chen; He, Cheng; Liu, Xiao-Ping; Lu, Ming-Hui; Zhu, Shi-Ning; Chen, Yan-Feng
2017-09-01
The topological phase of matter, originally proposed and first demonstrated in fermionic electronic systems, has drawn considerable research attention in the past decades due to its robust transport of edge states and its potential with respect to future quantum information, communication, and computation. Recently, searching for such a unique material phase in bosonic systems has become a hot research topic worldwide. So far, many bosonic topological models and methods for realizing them have been discovered in photonic systems, acoustic systems, mechanical systems, etc. These discoveries have certainly yielded vast opportunities in designing material phases and related properties in the topological domain. In this review, we first focus on some of the representative photonic topological models and employ the underlying Dirac model to analyze the edge states and geometric phase. On the basis of these models, three common types of two-dimensional topological photonic systems are discussed: 1) photonic quantum Hall effect with broken time-reversal symmetry; 2) photonic topological insulator and the associated pseudo-time-reversal symmetry-protected mechanism; 3) time/space periodically modulated photonic Floquet topological insulator. Finally, we provide a summary and extension of this emerging field, including a brief introduction to the Weyl point in three-dimensional systems.
Turbulent equipartitions in two dimensional drift convection
International Nuclear Information System (INIS)
Isichenko, M.B.; Yankov, V.V.
1995-01-01
Unlike the thermodynamic equipartition of energy in conservative systems, turbulent equipartitions (TEP) describe strongly non-equilibrium systems such as turbulent plasmas. In turbulent systems, energy is no longer a good invariant, but one can utilize the conservation of other quantities, such as adiabatic invariants, frozen-in magnetic flux, entropy, or combination thereof, in order to derive new, turbulent quasi-equilibria. These TEP equilibria assume various forms, but in general they sustain spatially inhomogeneous distributions of the usual thermodynamic quantities such as density or temperature. This mechanism explains the effects of particle and energy pinch in tokamaks. The analysis of the relaxed states caused by turbulent mixing is based on the existence of Lagrangian invariants (quantities constant along fluid-particle or other orbits). A turbulent equipartition corresponds to the spatially uniform distribution of relevant Lagrangian invariants. The existence of such turbulent equilibria is demonstrated in the simple model of two dimensional electrostatically turbulent plasma in an inhomogeneous magnetic field. The turbulence is prescribed, and the turbulent transport is assumed to be much stronger than the classical collisional transport. The simplicity of the model makes it possible to derive the equations describing the relaxation to the TEP state in several limits
Buckled two-dimensional Xene sheets.
Molle, Alessandro; Goldberger, Joshua; Houssa, Michel; Xu, Yong; Zhang, Shou-Cheng; Akinwande, Deji
2017-02-01
Silicene, germanene and stanene are part of a monoelemental class of two-dimensional (2D) crystals termed 2D-Xenes (X = Si, Ge, Sn and so on) which, together with their ligand-functionalized derivatives referred to as Xanes, are comprised of group IVA atoms arranged in a honeycomb lattice - similar to graphene but with varying degrees of buckling. Their electronic structure ranges from trivial insulators, to semiconductors with tunable gaps, to semi-metallic, depending on the substrate, chemical functionalization and strain. More than a dozen different topological insulator states are predicted to emerge, including the quantum spin Hall state at room temperature, which, if realized, would enable new classes of nanoelectronic and spintronic devices, such as the topological field-effect transistor. The electronic structure can be tuned, for example, by changing the group IVA element, the degree of spin-orbit coupling, the functionalization chemistry or the substrate, making the 2D-Xene systems promising multifunctional 2D materials for nanotechnology. This Perspective highlights the current state of the art and future opportunities in the manipulation and stability of these materials, their functions and applications, and novel device concepts.
Experimental Methods Related to Coupled Fast-Thermal Systems at the RB Reactor
International Nuclear Information System (INIS)
Pesic, M.
2002-01-01
In addition to the review of RB reactor characteristics this presentation is focused on the coupled fast-thermal systems achieved at the reactor. The following experimental methods are presented: neutron spectra measurements; steady state experiments and kinetic measurements ( β eff ) related to the coupled fast-thermal cores
A cermet fuel reactor for nuclear thermal propulsion
Kruger, Gordon
1991-01-01
Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.
A cermet fuel reactor for nuclear thermal propulsion
International Nuclear Information System (INIS)
Kruger, G.
1991-01-01
Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk
Thermal hydraulic stability in a pressure tube nuclear reactor
International Nuclear Information System (INIS)
Villani, A.; Ravetta, R.; Mansani, L.
1986-01-01
The CIRENE plant which will undergo preoperational tests in the near future is equipped with a 40 MW(e) Heavy Water moderated Boiling Light Water cooled Reactor (HWBLWR); at the start-up and up to about 30 % of nominal power, the necessary low coolant density is obtained injecting into the core a mixture of liquid and steam. To verify the thermal-hydraulic stability of the plant in this situation, tests have been carried out in a facility simulating two full scale power channels; the system stability has been confirmed in the reference conditions, and is not reduced by even a significant reduction of the liquid flowrate, where a decrease in liquid temperature has some negative effect and steam flowrate has a small influence. (author)
Rapid thermal transient in a reactor coolant channel
International Nuclear Information System (INIS)
Cherubini, A.
1986-01-01
This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow
Integral test of JENDL-3.3 for thermal reactors
International Nuclear Information System (INIS)
Okumura, Keisuke; Mori, Takamasa
2003-01-01
Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)
Periphyton crops and productivity in a reactor thermal effluent
International Nuclear Information System (INIS)
Tilly, L.J.
1975-01-01
Samples of periphyton grown for two weeks on microscope slides in surface waters of the reactor cooling reservoir, Par Pond, were examined for differences in species composition, diversity, standing crop, and 14 C uptake relatable to 7 positions in the thermal effluent. For stations which differed in average temperature by less than 5 0 C, weight specific productivity differed by a factor of 7. Periphyton biomass differed more than fivefold between stations 5.5 0 C apart. For most incubation intervals, both weight specific productivity and accumulated crop correlated highly with the average growing temperature, but slopes of regressions from consecutive periods often differed greatly while species composition and temperauture regime changed only slightly. Recent experiments indicate that observed differences may be due to interactions between nutrients and temperatures. (U.S.)
Probabilistic structural integrity of reactor vessel under pressurized thermal shock
International Nuclear Information System (INIS)
Myung Jo Hhung; Young Hwan Choi; Hho Jung Kim; Changheui Jang
2005-01-01
Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel. A round robin consisting of 1 prerequisite study and 5 cases for probabilistic approaches is proposed, and all organizations interested are invited. The problems are solved and their results are compared to issue some recommendation of best practices in this area and to assure an understanding of the key parameters of this type of approach, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria. Six participants from 3 organizations in Korea responded to the problem and their results are compiled in this study. (authors)
Phytoplankton distribution in three thermally distinct reactor cooling reservoirs
International Nuclear Information System (INIS)
Wilde, E.W.
1983-01-01
Phytoplankton community structure was studied in relation to physicochemical characteristics of three South Carolina reservoirs in close proximity and of similar age and bottom type. Thermal alteration, resulting from the input of cooling water from a nuclear reactor, was substantially different in each reservoir. This provided an opportunity to compare water temperature effects separated from season. Water temperature (when examined independently in statistical models) appeared to be less important than other environmental variables in determining phytoplankton community structure. Pond C, a reservoir receiving intensely heated effluent (> 20 0 C ΔT), displayed low species diversity (Shannon-Weaver H 0 C in summer. Par Pond, having a maximum ΔT of 5 0 C, displayed no temperature-induced alteration of phytoplankton community structure
Thermal radiation in gas core nuclear reactors for space propulsion
International Nuclear Information System (INIS)
Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.
1994-01-01
A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs
Statistical thermodynamics of a two-dimensional relativistic gas.
Montakhab, Afshin; Ghodrat, Malihe; Barati, Mahmood
2009-03-01
In this paper we study a fully relativistic model of a two-dimensional hard-disk gas. This model avoids the general problems associated with relativistic particle collisions and is therefore an ideal system to study relativistic effects in statistical thermodynamics. We study this model using molecular-dynamics simulation, concentrating on the velocity distribution functions. We obtain results for x and y components of velocity in the rest frame (Gamma) as well as the moving frame (Gamma;{'}) . Our results confirm that Jüttner distribution is the correct generalization of Maxwell-Boltzmann distribution. We obtain the same "temperature" parameter beta for both frames consistent with a recent study of a limited one-dimensional model. We also address the controversial topic of temperature transformation. We show that while local thermal equilibrium holds in the moving frame, relying on statistical methods such as distribution functions or equipartition theorem are ultimately inconclusive in deciding on a correct temperature transformation law (if any).
Thermal hydraulics model for Sandia's annular core research reactor
International Nuclear Information System (INIS)
Rao, Dasari V.; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.
1988-01-01
A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within ± 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)
Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA
International Nuclear Information System (INIS)
Ninokata, Hisashi
2012-01-01
The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)
The Storage of Thermal Reactor Safety Analysis data (STRESA)
International Nuclear Information System (INIS)
Tanarro Colodron, J.
2016-01-01
Full text: Storage of Thermal Reactor Safety Analysis data (STRESA) is an online information system that contains three technical databases: 1) European Nuclear Research Facilities, open to all online visitors; 2) Nuclear Experiments, available only to registered users; 3) Results Data, being the core content of the information system, its availability depends on the role and organisation of each user. Its main purpose is to facilitate the exchange of experimental data produced by large Euratom funded scientific projects addressing severe accidents, providing at the same time a secure repository for this information. Due to its purpose and architecture, it has become an important asset for networks of excellence as SARNET or NUGENIA. The Severe Accident ResearchNetwork of Excellence (SARNET)was set up in 2004 under the aegis of the research Euratom Framework Programmes to study severe accidents in watercooled nuclear power plants. Coordinated by the IRSN, SARNET unites 43 organizations involved in research on nuclear reactor safety in 18 European countries plus the USA, Canada, South Korea and India. In 2013, SARNET became fully integrated in the Technical Area N2(TA2), named “Severe accidents” of NUGENIA association, devoted to R&D on fission technology of Generation II and III. (author
Unsteady thermal analysis of gas-cooled fast reactor core
International Nuclear Information System (INIS)
Lakkis, I.A.
1993-01-01
This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously
Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'
Energy Technology Data Exchange (ETDEWEB)
Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)
1982-01-01
The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.
International Nuclear Information System (INIS)
Andrieux, Chantal
1976-03-01
The neutronic evolution of the reacteur Sena during the first and second cycles is presented. The experimental power distributions, obtained from the in-core instrumentation evaluation code CIRCE are compared with the theoretical powers calculated with the two-dimensional diffusion-evolution code EVOE. The CIRCE code allows: the study of the evolution of the principal parameters of the core, the comparison of the results of measured and theoretical estimates. Therefore this study has a great interest for the knowledge of the neutronic evolution of the core, as well as the validation of the refinement of theoretical estimation methods. The core calculation methods and requisite data for the evaluation of the measurements are presented after a brief description of the SENA core and its inner instrumentation. The principle of the in-core instrumentation evaluation code CIRCE, and calculation of the experimental power distributions and nuclear core parameters are then exposed. The results of the evaluation are discussed, with a comparison of the theoretical and experimental results. Taking account of the approximations used, these results, as far as the first and second cycles at SENA are concerned, are satisfactory, the deviations between theoretical and experimental power distributions being lower than 3% at the middle of the reactor and 9% at the periphery [fr
Application of the REMIX thermal mixing calculation program for the Loviisa reactor
International Nuclear Information System (INIS)
Kokkonen, I.; Tuomisto, H.
1987-08-01
The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant
Thermal-hydraulic modeling needs for passive reactors
International Nuclear Information System (INIS)
Kelly, J.M.
1997-01-01
The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken
Thermal-hydraulic modeling needs for passive reactors
Energy Technology Data Exchange (ETDEWEB)
Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)
1997-07-01
The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.
Testing ENDF/B-V data for thermal reactors
International Nuclear Information System (INIS)
Craig, D.S.
1982-10-01
Lattice parameters have been calculated for some thermal reactor benchmark lattices using ENDF/B-V data. These lattices were TRX-1, -2; BAPL-UO 2 -1,-2,-3; BNL-ThO 2 - 233 UO 2 -H 2 0-1,-2,-3; MIT-4,-5,-6; and PNL-31,-33,-35 (infinite lattices). In addition, parameters were calculated for 3 ZEEP lattices, 3 High-Conversion U0 2 -H 2 0 lattices, and 7 BNL-Th0 2 - 233 U0 2 -D 2 0 lattices. These calculations were made using the integral transport cell code RAHAB with the resonance reaction rates obtained using the OZMA code operating in the discrete ordinate mode. This code calculates the resonance rates allowing for the interaction of all resonances. Four group reaction rates for use in method comparisons are given for several lattices. The author discusses the use of the OZMA code for these calculations, including the choice of options and the orders of the angular quadratures, and compares results obtained using the CRNL thermal scattering data with those obtained using ENDF/B data
Validation of containment thermal hydraulic computer codes for VVER reactor
Energy Technology Data Exchange (ETDEWEB)
Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)
2005-07-01
Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to
Thermal durability of modified Synroc material as reactor fuel matrix
International Nuclear Information System (INIS)
Kikuchi, Akira; Kanazawa, Hiroyuki; Togashi, Yoshihiro; Matumoto, Seiichiro; Nishino, Yasuharu; Ohwada, Isao; Nakata, Masahito; Amano, Hidetoshi; Mitamura, Hisayoshi
1994-08-01
A Synroc, a polyphase titanate ceramics composed of three mineral phases (perovskite, hollandite and zirconolite), has an excellent performance of immobilization of high level nuclear waste. A working group in the Department of Hot Laboratories paid special attention to this merit and started a development study on a LWR fuel named 'Waste Disposal Possible (WDP) Fuel', which has the two functions of a reactor fuel and a waste form. The present paper mainly describes thermal durability of a modified Synroc material, which is essentially important for applying the material to a fuel matrix. The two kinds of Synroc specimens, designated 'SM' as modified and 'SB' as a reference, were prepared by hot-pressing and annealed at 1200degC to 1500degC for 30 min in air. Unexpected and peculiar spherical voids were observed in the specimen SM at 1400degC and 1500degC, which caused the specimen swelling. The formation of the voids depends significantly on the existence of spherical precipitates seen in the as-fabricated specimen including latent micropores with high pressure. On the other hand, the heat treatment at 1500degC formed additional new phases, designated 'Phase A' for the specimen SB and 'Phase X' for SM. Phase A is a decomposition product of hollandite and Phase X a reaction product of Phase A and perovskite in the spherical voids. Furthermore, additional information and thermal properties examined are presented in Appendix 1 and Appendix 2, respectively. It was recognized that the modified Synroc specimen SM had excellent thermal properties. (author)
Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor
International Nuclear Information System (INIS)
Merola, M.; Biggio, M.
1989-01-01
Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems
Seismically constrained two-dimensional crustal thermal structure of ...
Indian Academy of Sciences (India)
oil bearing sedimentary basins of India is situated at ... related to the evolution of the western margin of India ... of approximately 46 mW/m2 for stable continental areas. This makes the region interesting from ... for hydrocarbon accumulation.
Two-dimensional vibrational-electronic spectroscopy
Courtney, Trevor L.; Fox, Zachary W.; Slenkamp, Karla M.; Khalil, Munira
2015-10-01
Two-dimensional vibrational-electronic (2D VE) spectroscopy is a femtosecond Fourier transform (FT) third-order nonlinear technique that creates a link between existing 2D FT spectroscopies in the vibrational and electronic regions of the spectrum. 2D VE spectroscopy enables a direct measurement of infrared (IR) and electronic dipole moment cross terms by utilizing mid-IR pump and optical probe fields that are resonant with vibrational and electronic transitions, respectively, in a sample of interest. We detail this newly developed 2D VE spectroscopy experiment and outline the information contained in a 2D VE spectrum. We then use this technique and its single-pump counterpart (1D VE) to probe the vibrational-electronic couplings between high frequency cyanide stretching vibrations (νCN) and either a ligand-to-metal charge transfer transition ([FeIII(CN)6]3- dissolved in formamide) or a metal-to-metal charge transfer (MMCT) transition ([(CN)5FeIICNRuIII(NH3)5]- dissolved in formamide). The 2D VE spectra of both molecules reveal peaks resulting from coupled high- and low-frequency vibrational modes to the charge transfer transition. The time-evolving amplitudes and positions of the peaks in the 2D VE spectra report on coherent and incoherent vibrational energy transfer dynamics among the coupled vibrational modes and the charge transfer transition. The selectivity of 2D VE spectroscopy to vibronic processes is evidenced from the selective coupling of specific νCN modes to the MMCT transition in the mixed valence complex. The lineshapes in 2D VE spectra report on the correlation of the frequency fluctuations between the coupled vibrational and electronic frequencies in the mixed valence complex which has a time scale of 1 ps. The details and results of this study confirm the versatility of 2D VE spectroscopy and its applicability to probe how vibrations modulate charge and energy transfer in a wide range of complex molecular, material, and biological systems.
Two-dimensional silica opens new perspectives
Büchner, Christin; Heyde, Markus
2017-12-01
In recent years, silica films have emerged as a novel class of two-dimensional (2D) materials. Several groups succeeded in epitaxial growth of ultrathin SiO2 layers using different growth methods and various substrates. The structures consist of tetrahedral [SiO4] building blocks in two mirror symmetrical planes, connected via oxygen bridges. This arrangement is called a silica bilayer as it is the thinnest 2D arrangement with the stoichiometry SiO2 known today. With all bonds saturated within the nano-sheet, the interaction with the substrate is based on van der Waals forces. Complex ring networks are observed, including hexagonal honeycomb lattices, point defects and domain boundaries, as well as amorphous domains. The network structures are highly tuneable through variation of the substrate, deposition parameters, cooling procedure, introducing dopants or intercalating small species. The amorphous networks and structural defects were resolved with atomic resolution microscopy and modeled with density functional theory and molecular dynamics. Such data contribute to our understanding of the formation and characteristic motifs of glassy systems. Growth studies and doping with other chemical elements reveal ways to tune ring sizes and defects as well as chemical reactivities. The pristine films have been utilized as molecular sieves and for confining molecules in nanocatalysis. Post growth hydroxylation can be used to tweak the reactivity as well. The electronic properties of silica bilayers are favourable for using silica as insulators in 2D material stacks. Due to the fully saturated atomic structure, the bilayer interacts weakly with the substrate and can be described as quasi-freestanding. Recently, a mm-scale film transfer under structure retention has been demonstrated. The chemical and mechanical stability of silica bilayers is very promising for technological applications in 2D heterostacks. Due to the impact of this bilayer system for glass science
Two-dimensional vibrational-electronic spectroscopy
Energy Technology Data Exchange (ETDEWEB)
Courtney, Trevor L.; Fox, Zachary W.; Slenkamp, Karla M.; Khalil, Munira, E-mail: mkhalil@uw.edu [Department of Chemistry, University of Washington, Box 351700, Seattle, Washington 98195 (United States)
2015-10-21
Two-dimensional vibrational-electronic (2D VE) spectroscopy is a femtosecond Fourier transform (FT) third-order nonlinear technique that creates a link between existing 2D FT spectroscopies in the vibrational and electronic regions of the spectrum. 2D VE spectroscopy enables a direct measurement of infrared (IR) and electronic dipole moment cross terms by utilizing mid-IR pump and optical probe fields that are resonant with vibrational and electronic transitions, respectively, in a sample of interest. We detail this newly developed 2D VE spectroscopy experiment and outline the information contained in a 2D VE spectrum. We then use this technique and its single-pump counterpart (1D VE) to probe the vibrational-electronic couplings between high frequency cyanide stretching vibrations (ν{sub CN}) and either a ligand-to-metal charge transfer transition ([Fe{sup III}(CN){sub 6}]{sup 3−} dissolved in formamide) or a metal-to-metal charge transfer (MMCT) transition ([(CN){sub 5}Fe{sup II}CNRu{sup III}(NH{sub 3}){sub 5}]{sup −} dissolved in formamide). The 2D VE spectra of both molecules reveal peaks resulting from coupled high- and low-frequency vibrational modes to the charge transfer transition. The time-evolving amplitudes and positions of the peaks in the 2D VE spectra report on coherent and incoherent vibrational energy transfer dynamics among the coupled vibrational modes and the charge transfer transition. The selectivity of 2D VE spectroscopy to vibronic processes is evidenced from the selective coupling of specific ν{sub CN} modes to the MMCT transition in the mixed valence complex. The lineshapes in 2D VE spectra report on the correlation of the frequency fluctuations between the coupled vibrational and electronic frequencies in the mixed valence complex which has a time scale of 1 ps. The details and results of this study confirm the versatility of 2D VE spectroscopy and its applicability to probe how vibrations modulate charge and energy transfer in a
Test program for NIS calibration to reactor thermal output in HTTR
International Nuclear Information System (INIS)
Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko
2000-03-01
Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)
International Nuclear Information System (INIS)
Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya
2006-10-01
Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (author)
Sol-gel process for thermal reactor fuel fabrication
International Nuclear Information System (INIS)
Mukerjee, S.K.
2008-01-01
Full text: Sol-gel processes have revolutionized conventional ceramic technology by providing extremely fine and uniform powders for the fabrication of ceramics. The use of this technology for nuclear fuel fabrication has also been explored in many countries. Unlike the conventional sol-gel process, sol-gel process for nuclear fuels tries to eliminate the preparation of powders in view of the toxic nature of the powders particularly those of plutonium and 233 U. The elimination of powder handling thus makes this process more readily amenable for use in glove boxes or for remote handling. In this process, the first step is the preparation of microspheres of the fuel material from a solution which is then followed by vibro-compaction of these microspheres of different sizes to obtain the required smear density of fuel inside a pin. The maximum achievable packing density of 92 % makes it suitable for fast reactors only. With a view to extend the applicability of sol-gel process for thermal reactor fuel fabrication the concept of converting the gel microspheres derived from sol-gel process, to the pellets, has been under investigation for several years. The unique feature of this process is that it combines the advantages of sol-gel process for the preparation of fuel oxide gel microspheres of reproducible quality with proven irradiation behavior of the pellet fuel. One of the important pre-requisite for the success of this process is the preparation of soft oxide gel microspheres suitable for conversion to dense pellets free from berry structure. Studies on the internal gelation process, one of the many variants of sol-gel process, for obtaining soft oxide gel microspheres suitable for gel pelletisation is now under investigation at BARC. Some of the recent findings related to Sol-Gel Microsphere Pelletisation (SGMP) in urania-plutonia and thoria-urania systems will be presented
Validation of thermal hydraulic codes for fusion reactors safety
International Nuclear Information System (INIS)
Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.
2006-01-01
A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)
VISA-2, Reactor Vessel Failure Probability Under Thermal Shock
International Nuclear Information System (INIS)
Simonen, F.; Johnson, K.
1992-01-01
1 - Description of program or function: VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjust on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. 2 - Method of solution: The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation. 3 - Restrictions on the complexity of the problem: Maxima of 30 welds. VISA2 models only the belt-line (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only
International Nuclear Information System (INIS)
Meuschke, R.E.; Boyd, C.H.
1989-01-01
This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor
Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation
International Nuclear Information System (INIS)
Grotz, S.; Ghoniem, N.M.
1986-02-01
The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)
Assessment of RELAP5-3D copyright using data from two-dimensional RPI flow tests
International Nuclear Information System (INIS)
Davis, C.B.
1998-01-01
The capability of the RELAP5-3D copyright computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code's logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved
Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations
International Nuclear Information System (INIS)
Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.
1979-01-01
The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements
Two Dimensional Heat Transfer around Penetrations in Multilayer Insulation
Johnson, Wesley L.; Kelly, Andrew O.; Jumper, Kevin M.
2012-01-01
The objective of this task was to quantify thermal losses involving integrating MLI into real life situations. Testing specifically focused on the effects of penetrations (including structural attachments, electrical conduit/feedthroughs, and fluid lines) through MLI. While there have been attempts at quantifying these losses both analytically and experimentally, none have included a thorough investigation of the methods and materials that could be used in such applications. To attempt to quantify the excess heat load coming into the system due to the integration losses, a calorimeter was designed to study two dimensional heat transfer through penetrated MLI. The test matrix was designed to take as many variables into account as was possible with the limited test duration and system size. The parameters varied were the attachment mechanism, the buffer material (for buffer attachment mechanisms only), the thickness of the buffer, and the penetration material. The work done under this task is an attempt to measure the parasitic heat loads and affected insulation areas produced by system integration, to model the parasitic loads, and from the model produce engineering equations to allow for the determination of parasitic heat loads in future applications. The methods of integration investigated were no integration, using a buffer to thermally isolate the strut from the MLI, and temperature matching the MLI on the strut. Several materials were investigated as a buffer material including aerogel blankets, aerogel bead packages, cryolite, and even an evacuated vacuum space (in essence a no buffer condition).
International Nuclear Information System (INIS)
Honda, T.; Maki, K.; Okazaki, T.
1994-01-01
Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs
Temperature variation of criticality of thermal reactor lattices
International Nuclear Information System (INIS)
Velner, S.; Rothenstein, W.
1975-01-01
Departures from the asymptotic mode in the experimental setup have been examined in detail for two assemblies, one exponential, the other critical. It was found that the flux shape differed noticeably from the asymptotic mode in the core region especially for the exponential assemblies. On the other hand the departure from the fundamental mode has very little effect on the change of material buckling with temperature. Results of the calculations and their comparison with experiment are presented. The variation of material buckling with temperature is the same for ENDF/B-II and for ENDF/B-IV data, both for asymptotic reactor theory and for the buckling values derived from the flux calculated with the SN code. The results obtained with ENDF/B-IV data for both lattices are shown. In the small exponential assembly the results derived from S-4 calculations are compared with experiment. In the critical assembly the ratio of U-238 to U-235 fissions delta 28 and the relative conversion ratio - the ratio of U-238 captures to U-235 fissions in the lattice compared with the same quantity in a thermal column - are also shown. In both cases the experimental change of buckling with temperature is smaller than the calculated change. (B.G.)
Evaluation of the effectiveness of a thermal hygienization reactor
Directory of Open Access Journals (Sweden)
Daniel Borski
2011-01-01
Full Text Available For reasons of limiting the spread of serious transmissible diseases, with regard to the requirement for reducing landfill of biodegradable waste (which may or contains animal by-products and thus presents a potential risk to human and animal health and with a focus on supporting its separate collection, there has been created a legal framework for processing and hygienization of materials containing animal by-products. For the above reasons new technologies are being developed and implemented. These technologies are able to ensure the processing of biological waste containing animal by-products. As a practical result of the effort to ensure the hygienization of biowaste, a hygienization unit of own design, which uses the thermal way of hygienization, is presented in this work. The general part of the work defines a legislative framework for the assignment and gives technical parameters and minimum requirements for conversion that hygienization unit should be able to perform, including the limits for digestion residues and compost.In the experimental section there are described operational tests which document the technological process of hygienization depending on the aeration of the contents of the reactor. Experiment III outlines the validation process which uses contamination by indicator organisms, including subsequent checking of their occurrence as well as processing of the results of experiments and evaluation of the process of hygienization.
Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.
2015-12-01
Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.
A coupled nuclear reactor thermal energy storage system for enhanced load following operation
International Nuclear Information System (INIS)
Alameri, Saeed A.; King, Jeffrey C.
2013-01-01
Nuclear power plants operate most economically at a constant power level, providing base load electric power. In an energy grid containing a high fraction of renewable power sources, nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling a nuclear reactor to a large thermal energy storage block will allow the reactor to better respond to variable power demands. In the system described in this paper, a Prismatic core Advanced High Temperature Reactor supplies constant power to a lithium chloride molten salt thermal energy storage block that provides thermal power as needed to a closed Brayton cycle energy conversion system. During normal operation, the thermal energy storage block stores thermal energy during the night for use in the times of peak demand during the day. In this case, the nuclear reactor stays at a constant thermal power level. After a loss of forced circulation, the reactor reaches a shut down state in less than half an hour and the average fuel, graphite and coolant temperatures remain well within the design limits over the duration of the transient, demonstrating the inherent safety of the coupled system. (author)
Lie algebra contractions on two-dimensional hyperboloid
International Nuclear Information System (INIS)
Pogosyan, G. S.; Yakhno, A.
2010-01-01
The Inoenue-Wigner contraction from the SO(2, 1) group to the Euclidean E(2) and E(1, 1) group is used to relate the separation of variables in Laplace-Beltrami (Helmholtz) equations for the four corresponding two-dimensional homogeneous spaces: two-dimensional hyperboloids and two-dimensional Euclidean and pseudo-Euclidean spaces. We show how the nine systems of coordinates on the two-dimensional hyperboloids contracted to the four systems of coordinates on E 2 and eight on E 1,1 . The text was submitted by the authors in English.
Pseudo-harmonics method: an application to thermal reactors
International Nuclear Information System (INIS)
Silva, F.C. da; Rotenberg, S.; Thome Filho, Z.D.
1985-10-01
Several applications of the Pseudo-Harmonics method are presented, aiming to calculate the neutron flux and the perturbed eigenvalue of a nuclear reactor, like PWR, with three enrichment regions as Angra-1 reactor. In the reference reactor, perturbations of several types as global as local were simulated. The results were compared with those from the direct calculation. (E.G.) [pt
Two dimensional layered materials: First-principle investigation
Tang, Youjian
Two-dimensional layered materials have emerged as a fascinating research area due to their unique physical and chemical properties, which differ from those of their bulk counterparts. Some of these unique properties are due to carriers and transport being confined to 2 dimensions, some are due to lattice symmetry, and some arise from their large surface area, gateability, stackability, high mobility, spin transport, or optical accessibility. How to modify the electronic and magnetic properties of two-dimensional layered materials for desirable long-term applications or fundamental physics is the main focus of this thesis. We explored the methods of adsorption, intercalation, and doping as ways to modify two-dimensional layered materials, using density functional theory as the main computational methodology. Chapter 1 gives a brief review of density functional theory. Due to the difficulty of solving the many-particle Schrodinger equation, density functional theory was developed to find the ground-state properties of many-electron systems through an examination of their charge density, rather than their wavefunction. This method has great application throughout the chemical and material sciences, such as modeling nano-scale systems, analyzing electronic, mechanical, thermal, optical and magnetic properties, and predicting reaction mechanisms. Graphene and transition metal dichalcogenides are arguably the two most important two-dimensional layered materials in terms of the scope and interest of their physical properties. Thus they are the main focus of this thesis. In chapter 2, the structure and electronic properties of graphene and transition metal dichalcogenides are described. Alkali adsorption onto the surface of bulk graphite and metal intecalation into transition metal dichalcogenides -- two methods of modifying properties through the introduction of metallic atoms into layered systems -- are described in chapter 2. Chapter 3 presents a new method of tuning
International Nuclear Information System (INIS)
Pazsit, I.; Analytis, G.T.
1980-01-01
In order to develop a method for monitoring control rod vibrations by neutron noise measurements, the noise induced by two-dimensional vibrations of control elements is investigated. The two-dimensional Green's function relating the small stochastic cross-section fluctuations to the neutron noise is determined for a rectangular slab reactor in the modified one-group theory, and subsequently, the neutron response to two-dimensional vibrating noise sources is investigated. Two possible diagnostical applications are considered: (a) the reconstruction of the mechanical trajectory of the vibrating element by neutron noise measurements, and (b) the possibility of locating the vibrating element in the core. (author)
International Nuclear Information System (INIS)
Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa
1986-01-01
Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)
Review of PSI studies on reactor physics and thermal fluid dynamics of pebble bed reactors
International Nuclear Information System (INIS)
Prasser, Horst-Michael
2014-01-01
Switzerland is member of the Generation IV International Forum (GIF). The related work takes entirely place at PSI in the working groups of Gas-Cooled Fast Reactors and Very High Temperature Reactors. In the past, PSI has performed experimental and theoretical studies on criticality issues of pebble beds at the PROTEUS reactor, as well as a preliminary risk assessment of a prototypal HTR as an input for a comparison of energy supply options. PROTEUS was a critical assembly with an annular driver zone. The central region was filled by arrangements of fuel spheres. The reactivity effect of a water ingress was investigated by simulating the water by polyethylene rods of different diameter inserted into the gaps of a regular package. For sub-criticality measurements in pebble beds, a built-in pulsed neutron source was used. The experimental results were used to validate diffusion and higher order neutron transport models. Concerning thermal hydraulics of gas flows, the vast experience of PSI is focused on hydrogen transport, accumulation, and dispersion in containments of light water reactors. The phenomena are comparable in many aspects to the fluid dynamic issues relevant to HTR. Experiments on hydrogen flows are performed for numerous scenarios in the large-scale containment test facility PANDA. Hydrogen is substituted by helium as a model fluid. An important generic aspect is turbulent mixing in the presence of strong stratification, which is relevant for HTR as well. In a parallel project, generic small-scale mixing experiments with a high density ratio of 1:7 are carried out in a horizontal rectangular channel, where helium and nitrogen flows are brought into contact downstream of the rear edge of a splitter plate. Due to the high density ratio, turbulent mixing is affected by strong non-Boussinesq effects. The measurements taken by Particle Imaging Velocimetry (PIV) and Laser Induced Fluorescence techniques are compared to RANS and LES simulations. Similar large
Two-dimensional horizontal model seismic test and analysis for HTGR core
International Nuclear Information System (INIS)
Ikushima, Takeshi; Honma, Toshiaki.
1988-05-01
The resistance against earthquakes of high-temperature gas-cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. The paper presented the test results of seismic behavior of a half scale two-dimensional horizontal slice core model and analysis. The following is a summary of the more important results. (1) When the core is subjected to the single axis excitation and simultaneous two-axis excitations to the core across-corners, it has elliptical motion. The core stays lumped motion at the low excitation frequencies. (2) When the load is placed on side fixed reflector blocks from outside to the core center, the core displacement and reflector impact reaction force decrease. (3) The maximum displacement occurs at simultaneous two-axis excitations. The maximum displacement occurs at the single axis excitation to the core across-flats. (4) The results of two-dimensional horizontal slice core model was compared with the results of two-dimensional vertical one. It is clarified that the seismic response of actual core can be predicted from the results of two-dimensional vertical slice core model. (5) The maximum reflector impact reaction force for seismic waves was below 60 percent of that for sinusoidal waves. (6) Vibration behavior and impact response are in good agreement between test and analysis. (author)
Efficient cycles for carbon capture CLC power plants based on thermally balanced redox reactors
Iloeje, Chukwunwike; Zhao, Zhenlong; Ghoniem, Ahmed F.
2015-01-01
undergoing oxidation and reduction. An earlier study showed that this thermal coupling between the oxidation and reduction reactors increases the efficiency by up to 2% points when implemented in a regenerative Brayton cycle. The present study extends
Geodesics on a hot plate: an example of a two-dimensional curved space
International Nuclear Information System (INIS)
Erkal, Cahit
2006-01-01
The equation of the geodesics on a hot plate with a radially symmetric temperature profile is derived using the Lagrangian approach. Numerical solutions are presented with an eye towards (a) teaching two-dimensional curved space and the metric used to determine the geodesics (b) revealing some characteristics of two-dimensional curved spacetime and (c) providing insight into understanding the curved space which emerges in teaching relativity. In order to provide a deeper insight, we also present the analytical solutions and show that they represent circles whose characteristics depend on curvature of the space, conductivity and the coefficient of thermal expansion
Geodesics on a hot plate: an example of a two-dimensional curved space
Energy Technology Data Exchange (ETDEWEB)
Erkal, Cahit [Department of Geology, Geography, and Physics, University of Tennessee, Martin, TN 38238 (United States)
2006-07-01
The equation of the geodesics on a hot plate with a radially symmetric temperature profile is derived using the Lagrangian approach. Numerical solutions are presented with an eye towards (a) teaching two-dimensional curved space and the metric used to determine the geodesics (b) revealing some characteristics of two-dimensional curved spacetime and (c) providing insight into understanding the curved space which emerges in teaching relativity. In order to provide a deeper insight, we also present the analytical solutions and show that they represent circles whose characteristics depend on curvature of the space, conductivity and the coefficient of thermal expansion.
VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling
International Nuclear Information System (INIS)
Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.
1983-04-01
VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail
Thermal-hydraulic code selection for modular high temperature gas-cooled reactors
Energy Technology Data Exchange (ETDEWEB)
Komen, E M.J.; Bogaard, J.P.A. van den
1995-06-01
In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).
International Nuclear Information System (INIS)
Hollaway, W.R.
1991-08-01
If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issue through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW e IFR capacity for every three MW e Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years)
Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core
International Nuclear Information System (INIS)
D'Utra Bitelli, U.
1993-01-01
This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)
Beginning Introductory Physics with Two-Dimensional Motion
Huggins, Elisha
2009-01-01
During the session on "Introductory College Physics Textbooks" at the 2007 Summer Meeting of the AAPT, there was a brief discussion about whether introductory physics should begin with one-dimensional motion or two-dimensional motion. Here we present the case that by starting with two-dimensional motion, we are able to introduce a considerable…
Two-dimensional black holes and non-commutative spaces
International Nuclear Information System (INIS)
Sadeghi, J.
2008-01-01
We study the effects of non-commutative spaces on two-dimensional black hole. The event horizon of two-dimensional black hole is obtained in non-commutative space up to second order of perturbative calculations. A lower limit for the non-commutativity parameter is also obtained. The observer in that limit in contrast to commutative case see two horizon
Solution of the two-dimensional spectral factorization problem
Lawton, W. M.
1985-01-01
An approximation theorem is proven which solves a classic problem in two-dimensional (2-D) filter theory. The theorem shows that any continuous two-dimensional spectrum can be uniformly approximated by the squared modulus of a recursively stable finite trigonometric polynomial supported on a nonsymmetric half-plane.
Two-dimensional Navier-Stokes turbulence in bounded domains
Clercx, H.J.H.; van Heijst, G.J.F.
In this review we will discuss recent experimental and numerical results of quasi-two-dimensional decaying and forced Navier–Stokes turbulence in bounded domains. We will give a concise overview of developments in two-dimensional turbulence research, with emphasis on the progress made during the
Two-dimensional Navier-Stokes turbulence in bounded domains
Clercx, H.J.H.; Heijst, van G.J.F.
2009-01-01
In this review we will discuss recent experimental and numerical results of quasi-two-dimensional decaying and forced Navier–Stokes turbulence in bounded domains. We will give a concise overview of developments in two-dimensional turbulence research, with emphasis on the progress made during the
Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors
International Nuclear Information System (INIS)
Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.
1992-01-01
The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)
Institute of Scientific and Technical Information of China (English)
无
2010-01-01
The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.
Study of two-dimensional Debye clusters using Brownian motion
International Nuclear Information System (INIS)
Sheridan, T.E.; Theisen, W.L.
2006-01-01
A two-dimensional Debye cluster is a system of n identical particles confined in a parabolic well and interacting through a screened Coulomb (i.e., a Debye-Hueckel or Yukawa) potential with a Debye length λ. Experiments were performed for 27 clusters with n=3-63 particles (9 μm diam) in a capacitively coupled 9 W rf discharge at a neutral argon pressure of 13.6 mTorr. In the strong-coupling regime each particle exhibits small amplitude Brownian motion about its equilibrium position. These motions were projected onto the center-of-mass and breathing modes and Fourier analyzed to give resonance curves from which the mode frequencies, amplitudes, and damping rates were determined. The ratio of the breathing frequency to the center-of-mass frequency was compared with theory to self-consistently determine the Debye shielding parameter κ, Debye length λ, particle charge q, and mode temperatures. It is found that 1 < or approx. κ < or approx. 2, and κ decreases weakly with n. The particle charge averaged over all measurements is -14 200±200 e, and q decreases slightly with n. The two center-of-mass modes and the breathing mode are found to have the same temperature, indicating that the clusters are in thermal equilibrium with the neutral gas. The average cluster temperature is 399±5 K
Two-dimensional modeling of conduction-mode laser welding
International Nuclear Information System (INIS)
Russo, A.J.
1984-01-01
WELD2D is a two-dimensional finite difference computer program suitable for modeling the conduction-mode welding process when the molten weld pool motion can be neglected. The code is currently structured to treat butt-welded geometries in a plane normal to the beam motion so that dissimilar materials may be considered. The surface heat transfer models used in the code include a Gaussian beam or uniform laser source, and a free electron theory reflectance calculation. Temperature-dependent material parameters are used in the reflectance calculation. Measured cold reflection data are used to include surface roughness or oxide effects until melt occurs, after which the surface is assumed to be smooth and clean. Blackbody reradiation and a simple natural convection model are also included in the upper surface boundary condition. Either an implicit or explicit finite-difference representation of the heat conduction equation in an enthalpy form is solved at each time step. This enables phase transition energies to be easily and accurately incorporated into the formulation. Temperature-dependent 9second-order polynominal dependence) thermal conductivities are used in the conduction calculations. Constant values of specific heat are used for each material phase. At present, material properties for six metals are included in the code. These are: aluminium, nickel, steel, molybdenum, copper and silicon
Energy Technology Data Exchange (ETDEWEB)
Otra Otra, F; Leira Rey, G
1971-07-01
In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.
International Nuclear Information System (INIS)
Takeshi, Y.; Keisuke, K.
1983-01-01
The multigroup neutron diffusion equation for two-dimensional triangular geometry is solved by the finite Fourier transformation method. Using the zero-th-order equation of the integral equation derived by this method, simple algebraic expressions for the flux are derived and solved by the alternating direction implicit method. In sample calculations for a benchmark problem of a fast breeder reactor, it is shown that the present method gives good results with fewer mesh points than the usual finite difference method
Econometric modelling of certain nuclear power systems based on thermal and fast breeder reactors
International Nuclear Information System (INIS)
Pavelescu, M.; Pioaru, C.; Ursu, I.
1988-01-01
Certain known economic analysis models for a LMFBR fast breeder and CANDU thermal solitary reactors are presented, based on the concepts of discounting and levelization. These models are subsequently utilized as a basis for establishing an original model for the econometric analysis of certain thermal reactor systems or/and fast breeder reactors. Case studies are subsequently conducted with the systems: 1-CANDU, 2-LMFBR, 3-CANDU + LMFBR which enables us to draw certain interesting conclusions for a long range nuclear power policy. (author)
Simulation of Thermal-hydraulic Process in Reactor of HTR-PM
International Nuclear Information System (INIS)
Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle
2014-01-01
This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)
Benchmark tests of JENDL-3.2 for thermal and fast reactors
International Nuclear Information System (INIS)
Takano, Hideki
1995-01-01
Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)
Test and application of thermal neutron radiography facility at Xi'an pulsed reactor
Yang Jun; Zhao Xiang Feng; Wang Dao Hua
2002-01-01
A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter
Thermal insulation of the high-temperature helium-cooled reactors
International Nuclear Information System (INIS)
Kharlamov, A.G.; Grebennik, V.N.
1979-01-01
Unlike the well-known thermal insulation methods, development of high-temperature helium reactors (HTGR) raises quite new problems. To understand these problems, it is necessary to consider behaviour of thermal insulation inside the helium circuit of HTGR and requirements imposed on it. Substantiation of these requirements is given in the presented paper
Thermal and fast reactor benchmark testing of ENDF/B-6.4
International Nuclear Information System (INIS)
Liu Guisheng
1999-01-01
The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved
Optimizing separations in online comprehensive two-dimensional liquid chromatography.
Pirok, Bob W J; Gargano, Andrea F G; Schoenmakers, Peter J
2018-01-01
Online comprehensive two-dimensional liquid chromatography has become an attractive option for the analysis of complex nonvolatile samples found in various fields (e.g. environmental studies, food, life, and polymer sciences). Two-dimensional liquid chromatography complements the highly popular hyphenated systems that combine liquid chromatography with mass spectrometry. Two-dimensional liquid chromatography is also applied to the analysis of samples that are not compatible with mass spectrometry (e.g. high-molecular-weight polymers), providing important information on the distribution of the sample components along chemical dimensions (molecular weight, charge, lipophilicity, stereochemistry, etc.). Also, in comparison with conventional one-dimensional liquid chromatography, two-dimensional liquid chromatography provides a greater separation power (peak capacity). Because of the additional selectivity and higher peak capacity, the combination of two-dimensional liquid chromatography with mass spectrometry allows for simpler mixtures of compounds to be introduced in the ion source at any given time, improving quantitative analysis by reducing matrix effects. In this review, we summarize the rationale and principles of two-dimensional liquid chromatography experiments, describe advantages and disadvantages of combining different selectivities and discuss strategies to improve the quality of two-dimensional liquid chromatography separations. © 2017 The Authors. Journal of Separation Science published by WILEY-VCH Verlag GmbH & Co. KGaA.
Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)
International Nuclear Information System (INIS)
Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.
1994-08-01
We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)
TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies
International Nuclear Information System (INIS)
Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.
1983-01-01
TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies
Maximisation of the Doppler effect in thermal reactors
International Nuclear Information System (INIS)
Bende, E.E.
1998-03-01
Increase of the fuel temperature in a nuclear reactor leads, or can lead, to (1) A Doppler broadening of the resonances of the nuclides in the fuel; (2) An expansion of the fuel; and (3) A shift of the Maxwellian part of the spectrum to higher energies. These processes together introduce a certain amount of reactivity, which can be expressed in the so-called fuel temperature reactivity coefficient. The reactivity effect of the third process is very small, because the Maxwell spectrum is to a major extent determined by the moderator temperature. Moreover, the reactivity effect due to an expansion of the fuel is small too, for most thermal systems. When the second and third processes can be neglected, the fuel temperature reactivity effect is fully determined by the Doppler effect. The fuel temperature reactivity coefficient is then called the Doppler coefficient of reactivity. The Doppler broadening of the resonances causes an increase of resonance absorption, due to a decrease of self-shielding. The competition between resonance fission at the one hand and resonance capture at the other hand determines the sign and magnitude of the reactivity induced by an increase of the fuel temperature. In well-designed nuclear reactors the Doppler effect due to resonance capture by fertile nuclides exceeds the Doppler effect due to resonance fission, which implies that an increase of the fuel temperature causes a negative reactivity effect and a correspondingly negative Doppler coefficient. Since the Doppler effect is a prompt effect, occurring simultaneously with the dissipation of kinetic energy of the fission products into temperature, it is very important in the study of rapid power transients. In this report, the Doppler coefficient of reactivity is defined in chapter 2. Chapter 3 discusses the geometry of the unit-cell for which the calculations are performed and describes the fuel types that have been investigated. In chapter 4 the 'Doppler efficiency' is introduced and
Exploring two-dimensional electron gases with two-dimensional Fourier transform spectroscopy
Energy Technology Data Exchange (ETDEWEB)
Paul, J.; Dey, P.; Karaiskaj, D., E-mail: karaiskaj@usf.edu [Department of Physics, University of South Florida, 4202 East Fowler Ave., Tampa, Florida 33620 (United States); Tokumoto, T.; Hilton, D. J. [Department of Physics, University of Alabama at Birmingham, Birmingham, Alabama 35294 (United States); Reno, J. L. [CINT, Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States)
2014-10-07
The dephasing of the Fermi edge singularity excitations in two modulation doped single quantum wells of 12 nm and 18 nm thickness and in-well carrier concentration of ∼4 × 10{sup 11} cm{sup −2} was carefully measured using spectrally resolved four-wave mixing (FWM) and two-dimensional Fourier transform (2DFT) spectroscopy. Although the absorption at the Fermi edge is broad at this doping level, the spectrally resolved FWM shows narrow resonances. Two peaks are observed separated by the heavy hole/light hole energy splitting. Temperature dependent “rephasing” (S{sub 1}) 2DFT spectra show a rapid linear increase of the homogeneous linewidth with temperature. The dephasing rate increases faster with temperature in the narrower 12 nm quantum well, likely due to an increased carrier-phonon scattering rate. The S{sub 1} 2DFT spectra were measured using co-linear, cross-linear, and co-circular polarizations. Distinct 2DFT lineshapes were observed for co-linear and cross-linear polarizations, suggesting the existence of polarization dependent contributions. The “two-quantum coherence” (S{sub 3}) 2DFT spectra for the 12 nm quantum well show a single peak for both co-linear and co-circular polarizations.
Energy Technology Data Exchange (ETDEWEB)
Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)
2013-07-01
A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics
International Nuclear Information System (INIS)
Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.
2013-01-01
A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics
KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo
International Nuclear Information System (INIS)
Cupini, E.; De Matteis, A.; Simonini, R.
1980-01-01
1 - Description of problem or function: KIM (K-infinite Monte Carlo) is a program which solves the steady-state linear transport equation for a fixed-source problem (or, by successive fixed-source runs, for the eigenvalue problem) in a two-dimensional infinite thermal reactor lattice. The main quantities computed in some broad energy groups are the following: - Fluxes and cross sections averaged over the region (i.e. a space portion that can be unconnected but contains everywhere the same homogeneous material), grouping of regions, the whole element. - Average absorption and fission rates per nuclide. - Average flux, absorption and production distributions versus energy. 2 - Method of solution: Monte Carlo simulation is used by tracing particle histories from fission birth down through the resonance region until absorption in the thermal range. The program is organised in three sections for fast, epithermal and thermal simulation, respectively; each section implements a particular model for both numerical techniques and cross section representation (energy groups in the fast section, groups or resonance parameters in the epithermal section, points in the thermal section). During slowing down (energy above 1 eV) nuclei are considered as stationary, with the exception of some resonance nuclei whose spacing between resonances is much greater than the resonance width. The Doppler broadening of s-wave resonances of these nuclides is taken into account by computing cross sections at the current neutron energy and at the temperature of the nucleus hit. During thermalization (energy below 1 eV) the thermal motion of some nuclides is also considered, by exploiting scattering kernels provided by the library for light water, heavy water and oxygen at several temperatures. KIM includes splitting and Russian roulette. A characteristic feature of the program is its approach to the lattice geometry. In fact, besides the usual continuous treatment of the geometry using the well
International Nuclear Information System (INIS)
Oh, Se-Jin; Kim, Young-Chul; Chung, Chin-Wook
2011-01-01
An interpolation algorithm for the evaluation of the spatial profile of plasma densities in a cylindrical reactor was developed for low gas pressures. The algorithm is based on a collisionless two-dimensional fluid model. Contrary to the collisional case, i.e., diffusion fluid model, the fitting algorithm depends on the aspect ratio of the cylindrical reactor. The spatial density profile of the collisionless fitting algorithm is presented in two-dimensional images and compared with the results of the diffusion fluid model.
Response of a thermal barrier system to acoustic excitation in a gas turbine nuclear reactor
International Nuclear Information System (INIS)
Betts, W.S. Jr.; Blevins, R.D.
1980-11-01
A gas turbine located within a High-Temperature Gas-Cooled Reactor (HTGR) induces high acoustic sound pressure levels into the primary coolant (helium). This acoustic loading induces high cycle fatigue stresses which may control the design of the thermal barrier system. This study examines the dynamic response of a thermal barrier configuration consisting of a fibrous insulation compressed against the reactor vessel by a coverplate which is held in position by a central attachment fixture. The results of dynamic vibration analyses indicate the effect of the plate size and curvature and the attachment size on the response of the thermal barrier
A method for statistical steady state thermal analysis of reactor cores
International Nuclear Information System (INIS)
Whetton, P.A.
1981-01-01
In a previous publication the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analysis are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations. (orig.)
ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor
Energy Technology Data Exchange (ETDEWEB)
Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)
2015-08-15
Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.
Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor
International Nuclear Information System (INIS)
Ustun, G.; Durmayaz, A.
2002-01-01
Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor
Functional inks and printing of two-dimensional materials.
Hu, Guohua; Kang, Joohoon; Ng, Leonard W T; Zhu, Xiaoxi; Howe, Richard C T; Jones, Christopher G; Hersam, Mark C; Hasan, Tawfique
2018-05-08
Graphene and related two-dimensional materials provide an ideal platform for next generation disruptive technologies and applications. Exploiting these solution-processed two-dimensional materials in printing can accelerate this development by allowing additive patterning on both rigid and conformable substrates for flexible device design and large-scale, high-speed, cost-effective manufacturing. In this review, we summarise the current progress on ink formulation of two-dimensional materials and the printable applications enabled by them. We also present our perspectives on their research and technological future prospects.
Third sound in one and two dimensional modulated structures
International Nuclear Information System (INIS)
Komuro, T.; Kawashima, H., Shirahama, K.; Kono, K.
1996-01-01
An experimental technique is developed to study acoustic transmission in one and two dimensional modulated structures by employing third sound of a superfluid helium film. In particular, the Penrose lattice, which is a two dimensional quasiperiodic structure, is studied. In two dimensions, the scattering of third sound is weaker than in one dimension. Nevertheless, the authors find that the transmission spectrum in the Penrose lattice, which is a two dimensional prototype of the quasicrystal, is observable if the helium film thickness is chosen around 5 atomic layers. The transmission spectra in the Penrose lattice are explained in terms of dynamical theory of diffraction
ONE-DIMENSIONAL AND TWO-DIMENSIONAL LEADERSHIP STYLES
Directory of Open Access Journals (Sweden)
Nikola Stefanović
2007-06-01
Full Text Available In order to motivate their group members to perform certain tasks, leaders use different leadership styles. These styles are based on leaders' backgrounds, knowledge, values, experiences, and expectations. The one-dimensional styles, used by many world leaders, are autocratic and democratic styles. These styles lie on the two opposite sides of the leadership spectrum. In order to precisely define the leadership styles on the spectrum between the autocratic leadership style and the democratic leadership style, leadership theory researchers use two dimensional matrices. The two-dimensional matrices define leadership styles on the basis of different parameters. By using these parameters, one can identify two-dimensional styles.
Wang, Yongjiang; Niu, Wenjuan; Ai, Ping
2016-12-01
Dynamic estimation of heat transfer through composting reactor wall was crucial for insulating design and maintaining a sanitary temperature. A model, incorporating conductive, convective and radiative heat transfer mechanisms, was developed in this paper to provide thermal resistance calculations for composting reactor wall. The mechanism of thermal transfer from compost to inner surface of structural layer, as a first step of heat loss, was important for improving insulation performance, which was divided into conduction and convection and discussed specifically in this study. It was found decreasing conductive resistance was responsible for the drop of insulation between compost and reactor wall. Increasing compost porosity or manufacturing a curved surface, decreasing the contact area of compost and the reactor wall, might improve the insulation performance. Upon modeling of heat transfers from compost to ambient environment, the study yielded a condensed and simplified model that could be used to conduct thermal resistance analysis for composting reactor. With theoretical derivations and a case application, the model was applicable for both dynamic estimation and typical composting scenario. Copyright © 2016 Elsevier Ltd. All rights reserved.
Parametric study on thermal-hydraulic characteristics of high conversion light water reactor
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Fujii, Sadao.
1988-11-01
To assess the feasibility of high conversion light water reactors (HCLWRs) from the thermal-hydraulic viewpoint, parametric study on thermal-hydraulic characteristics of HCLWR has been carried out by using a unit cell model. It is assumed that a HCLWR core is contained in a current 1000 MWe PWR plant. At the present study, reactor core parameters such as fuel pin diameter, pitch, core height and linear heat rate are widely and parametrically changed to survey the relation between these parameters and the basic thermal-hydraulic characteristics, i.e. maximum fuel temperature, minimum DNBR, reduction of reactor thermal output and so on. The validity of the unit cell model used has been ensured by comparison with the result of a subchannel analysis carried out for a whole core. (author)
Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors
Energy Technology Data Exchange (ETDEWEB)
Bodey, Isaac T [ORNL
2014-05-01
Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts
Measuring technique of super high temperature thermal properties of reactor core materials
International Nuclear Information System (INIS)
Ono, Akira; Baba, Tetsuya; Watanabe, Hideo; Matsumoto, Tsuyoshi
1998-01-01
In this study, thermal properties of reactor core materials used for water cooled reactors and FBR were tried to develop a technique to measure their melt states at less than 3,000degC in order to contribute more correct evaluation of the reactor core behavior at severe accident. Then, a thermal property measuring method of high temperature melt by using floating method was investigated and its fundamental design was begun to investigate under a base of optimum judgement on the air flow floating throw-down method. And, in order to measure emissivity of melt specimen surface essential for correct temperature measurement using the throw down method, a spectroscopic emissivity measuring unit using an ellipsometer was prepared and induced. On the thermal properties measurement using the holding method, a specimen container to measure thermal diffusiveness of the high temperature melts by using laser flashing method was tried to prepare. (G.K.)
Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion
Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei
2004-02-01
A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.
An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code
Energy Technology Data Exchange (ETDEWEB)
Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)
2017-01-15
Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.
Relevant thermal hydraulic aspects of advanced reactors design: status report
International Nuclear Information System (INIS)
1996-11-01
This status report provides an overview on the relevant thermalhydraulic aspects of advanced reactor designs (e.g. ABWR, AP600, SBWR, EPR, ABB 80+, PIUS, etc.). Since all of the advanced reactor concepts are at the design stage, the information and data available in the open literature are still very limited. Some characteristics of advanced reactor designs are provided together with selected phenomena identification and ranking tables. Specific needs for thermalhydraulic codes together with the list of relevant and important thermalhydraulic phenomena for advanced reactor designs are summarized with the purpose of providing some guidance in development of research plans for considering further code development and assessment needs and for the planning of experimental programs
Gas cooled thermal reactors with high temperatures (VHTR)
International Nuclear Information System (INIS)
Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.; Vasile, A.
2014-01-01
VHTR is one of the 6 concepts retained for the 4. generation of nuclear reactors, it is an upgraded version of the HTR-type reactor (High Temperature Reactors). 5 HTR reactors were operated in the world in the eighties, now 2 experimental HTR are working in China and Japan and 2 HTR with an output power of 100 MWe are being built in China. The purpose of the VHTR is to provide an helium at very high temperatures around 1000 Celsius degrees that could be used directly in a thermochemical way to produce hydrogen for instance. HTR reactors are interesting in terms of safety but it does not optimise the consumption of uranium and the production of wastes. This article presents a brief historical account of HTR-type reactors and their main design and safety features. The possibility of using HTR to burn plutonium is also presented as well as the possibility of closing the fuel cycle and of using thorium-uranium fuel. (A.C.)
Multisoliton formula for completely integrable two-dimensional systems
International Nuclear Information System (INIS)
Chudnovsky, D.V.; Chudnovsky, G.V.
1979-01-01
For general two-dimensional completely integrable systems, the exact formulae for multisoliton type solutions are given. The formulae are obtained algebrically from solutions of two linear partial differential equations
Two-dimensional electronic femtosecond stimulated Raman spectroscopy
Directory of Open Access Journals (Sweden)
Ogilvie J.P.
2013-03-01
Full Text Available We report two-dimensional electronic spectroscopy with a femtosecond stimulated Raman scattering probe. The method reveals correlations between excitation energy and excited state vibrational structure following photoexcitation. We demonstrate the method in rhodamine 6G.
Generalized similarity method in unsteady two-dimensional MHD ...
African Journals Online (AJOL)
user
International Journal of Engineering, Science and Technology. Vol. 1, No. 1, 2009 ... temperature two-dimensional MHD laminar boundary layer of incompressible fluid. ...... Φ η is Blasius solution for stationary boundary layer on the plate,. ( ). 0.
International Nuclear Information System (INIS)
Langenstein, M.; Streit, S.; Laipple, B.; Eitschberger, H.
2005-01-01
The determination of the thermal reactor power is traditionally be done by heat balance: 1) for a boiling water reactor (BWR) at the interface of reactor control volume and heat cycle. 2) for a pressurised-water reactor (PWR) at the interface of the steam generator control volume and turbine island on the secondary side. The uncertainty of these traditional methods is not easy to determine and can be in the range of several percent. Technical and legal regulations (e.g. 10CFR50) cover an estimated error of instrumentation up to 2% by increasing the design thermal reactor power for emergency analysis to 102 % of the licensed thermal reactor power. Basically the licensee has the duty to warrant at any time operation inside the analyzed region for thermal reactor power. This is normally done by keeping the indicated reactor power at the licensed 100% value. The better way is to use a method which allows a continuous warranty evaluation. The quantification of the level of fulfilment of this warranty is only achievable by a method which: 1) is independent of single measurements accuracies. 2) results in a certified quality of single process values and for the total heat cycle analysis. 3)leads to complete results including 2-sigma deviation especially for thermal reactor power. Here this method, which is called 'process data reconciliation based on VDI 2048 guideline', is presented [1, 2]. This method allows to determine the true process parameters with a statistical probability of 95%, by considering closed material, mass- and energy balances following the Gaussian correction principle. The amount of redundant process information and complexity of the process improves the final results. This represents the most probable state of the process with minimized uncertainty according to VDI 2048. Hence, calibration and control of the thermal reactor power are possible with low effort but high accuracy and independent of single measurement accuracies. Further more, VDI 2048
International Nuclear Information System (INIS)
Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki
1999-07-01
A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)
Topological aspect of disclinations in two-dimensional crystals
International Nuclear Information System (INIS)
Wei-Kai, Qi; Tao, Zhu; Yong, Chen; Ji-Rong, Ren
2009-01-01
By using topological current theory, this paper studies the inner topological structure of disclinations during the melting of two-dimensional systems. From two-dimensional elasticity theory, it finds that there are topological currents for topological defects in homogeneous equation. The evolution of disclinations is studied, and the branch conditions for generating, annihilating, crossing, splitting and merging of disclinations are given. (the physics of elementary particles and fields)
Structures of two-dimensional three-body systems
International Nuclear Information System (INIS)
Ruan, W.Y.; Liu, Y.Y.; Bao, C.G.
1996-01-01
Features of the structure of L = 0 states of a two-dimensional three-body model system have been investigated. Three types of permutation symmetry of the spatial part, namely symmetric, antisymmetric, and mixed, have been considered. A comparison has been made between the two-dimensional system and the corresponding three-dimensional one. The effect of symmetry on microscopic structures is emphasized. (author)
Study on two-dimensional induced signal readout of MRPC
International Nuclear Information System (INIS)
Wu Yucheng; Yue Qian; Li Yuanjing; Ye Jin; Cheng Jianping; Wang Yi; Li Jin
2012-01-01
A kind of two-dimensional readout electrode structure for the induced signal readout of MRPC has been studied in both simulation and experiments. Several MRPC prototypes are produced and a series of test experiments have been done to compare with the result of simulation, in order to verify the simulation model. The experiment results are in good agreement with those of simulation. This method will be used to design the two-dimensional signal readout mode of MRPC in the future work.
Controlled Interactions between Two Dimensional Layered Inorganic Nanosheets and Polymers
2016-06-15
AFRL-AFOSR-JP-TR-2016-0071 Controlled Interactions between Two Dimensional Layered Inorganic Nanosheets and Polymers Cheolmin Park YONSEI UNIVERSITY...Interactions between Two Dimensional Layered Inorganic Nanosheets and Polymers 5a. CONTRACT NUMBER 5b. GRANT NUMBER FA2386-14-1-4054 5c. PROGRAM ELEMENT...prospects for a variety of emerging applications in a broad range of fields, such as electronics, energy conversion and storage, catalysis and polymer
Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes
International Nuclear Information System (INIS)
Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.
2003-01-01
The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power
The theory of critical phenomena in two-dimensional systems
International Nuclear Information System (INIS)
Olvera de la C, M.
1981-01-01
An exposition of the theory of critical phenomena in two-dimensional physical systems is presented. The first six chapters deal with the mean field theory of critical phenomena, scale invariance of the thermodynamic functions, Kadanoff's spin block construction, Wilson's renormalization group treatment of critical phenomena in configuration space, and the two-dimensional Ising model on a triangular lattice. The second part of this work is made of four chapters devoted to the application of the ideas expounded in the first part to the discussion of critical phenomena in superfluid films, two-dimensional crystals and the two-dimensional XY model of magnetic systems. Chapters seven to ten are devoted to the following subjects: analysis of long range order in one, two, and three-dimensional physical systems. Topological defects in the XY model, in superfluid films and in two-dimensional crystals. The Thouless-Kosterlitz iterated mean field theory of the dipole gas. The renormalization group treatment of the XY model, superfluid films and two-dimensional crystal. (author)
Two-dimensional multifractal cross-correlation analysis
International Nuclear Information System (INIS)
Xi, Caiping; Zhang, Shuning; Xiong, Gang; Zhao, Huichang; Yang, Yonghong
2017-01-01
Highlights: • We study the mathematical models of 2D-MFXPF, 2D-MFXDFA and 2D-MFXDMA. • Present the definition of the two-dimensional N 2 -partitioned multiplicative cascading process. • Do the comparative analysis of 2D-MC by 2D-MFXPF, 2D-MFXDFA and 2D-MFXDMA. • Provide a reference on the choice and parameter settings of these methods in practice. - Abstract: There are a number of situations in which several signals are simultaneously recorded in complex systems, which exhibit long-term power-law cross-correlations. This paper presents two-dimensional multifractal cross-correlation analysis based on the partition function (2D-MFXPF), two-dimensional multifractal cross-correlation analysis based on the detrended fluctuation analysis (2D-MFXDFA) and two-dimensional multifractal cross-correlation analysis based on the detrended moving average analysis (2D-MFXDMA). We apply these methods to pairs of two-dimensional multiplicative cascades (2D-MC) to do a comparative study. Then, we apply the two-dimensional multifractal cross-correlation analysis based on the detrended fluctuation analysis (2D-MFXDFA) to real images and unveil intriguing multifractality in the cross correlations of the material structures. At last, we give the main conclusions and provide a valuable reference on how to choose the multifractal algorithms in the potential applications in the field of SAR image classification and detection.
Two-Dimensional Materials for Sensing: Graphene and Beyond
Directory of Open Access Journals (Sweden)
Seba Sara Varghese
2015-09-01
Full Text Available Two-dimensional materials have attracted great scientific attention due to their unusual and fascinating properties for use in electronics, spintronics, photovoltaics, medicine, composites, etc. Graphene, transition metal dichalcogenides such as MoS2, phosphorene, etc., which belong to the family of two-dimensional materials, have shown great promise for gas sensing applications due to their high surface-to-volume ratio, low noise and sensitivity of electronic properties to the changes in the surroundings. Two-dimensional nanostructured semiconducting metal oxide based gas sensors have also been recognized as successful gas detection devices. This review aims to provide the latest advancements in the field of gas sensors based on various two-dimensional materials with the main focus on sensor performance metrics such as sensitivity, specificity, detection limit, response time, and reversibility. Both experimental and theoretical studies on the gas sensing properties of graphene and other two-dimensional materials beyond graphene are also discussed. The article concludes with the current challenges and future prospects for two-dimensional materials in gas sensor applications.
Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor
International Nuclear Information System (INIS)
Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C.; Palma, Daniel A.P.
2015-01-01
Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)
Two-dimensional model of a freely expanding plasma
International Nuclear Information System (INIS)
Khalid, Q.
1975-01-01
The free expansion of an initially confined plasma is studied by the computer experiment technique. The research is an extension to two dimensions of earlier work on the free expansion of a collisionless plasma in one dimension. In the two-dimensional rod model, developed in this research, the plasma particles, electrons and ions are modeled as infinitely long line charges or rods. The line charges move freely in two dimensions normal to their parallel axes, subject only to a self-consistent electric field. Two approximations, the grid approximation and the periodic boundary condition are made in order to reduce the computation time. In the grid approximation, the space occupied by the plasma at a given time is divided into boxes. The particles are subject to an average electric field calculated for that box assuming that the total charge within each box is located at the center of the box. However, the motion of each particle is exactly followed. The periodic boundary condition allows us to consider only one-fourth of the total number of particles of the plasma, representing the remaining three-fourths of the particles as symmetrically placed images of those whose positions are calculated. This approximation follows from the expected azimuthal symmetry of the plasma. The dynamics of the expansion are analyzed in terms of average ion and electron positions, average velocities, oscillation frequencies and relative distribution of energy between thermal, flow and electric field energies. Comparison is made with previous calculations of one-dimensional models which employed plane, spherical or cylindrical sheets as charged particles. In order to analyze the effect of the grid approximation, the model is solved for two different grid sizes and for each grid size the plasma dynamics is determined. For the initial phase of expansion, the agreement for the two grid sizes is found to be good
Proton and hydrogen transport through two-dimensional monolayers
International Nuclear Information System (INIS)
Seel, Max; Pandey, Ravindra
2016-01-01
Diffusion of protons and hydrogen atoms in representative two-dimensional materials is investigated. Specifically, density functional calculations were performed on graphene, hexagonal boron nitride (h-BN), phosphorene, silicene, and molybdenum disulfide (MoS 2 ) monolayers to study the surface interaction and penetration barriers for protons and hydrogen atoms employing finite cluster models. The calculated barrier heights correlate approximately with the size of the opening formed by the three-fold open sites in the monolayers considered. They range from 1.56 eV (proton) and 4.61 eV (H) for graphene to 0.12 eV (proton) and 0.20 eV (H) for silicene. The results indicate that only graphene and h-BN monolayers have the potential for membranes with high selective permeability. The MoS 2 monolayer behaves differently: protons and H atoms become trapped between the outer S layers in the Mo plane in a well with a depth of 1.56 eV (proton) and 1.5 eV (H atom), possibly explaining why no proton transport was detected, suggesting MoS 2 as a hydrogen storage material instead. For graphene and h-BN, off-center proton penetration reduces the barrier to 1.38 eV for graphene and 0.11 eV for h-BN. Furthermore, Pt acting as a substrate was found to have a negligible effect on the barrier height. In defective graphene, the smallest barrier for proton diffusion (1.05 eV) is found for an oxygen-terminated defect. Therefore, it seems more likely that thermal protons can penetrate a monolayer of h-BN but not graphene and defects are necessary to facilitate the proton transport in graphene. (paper)
Proton and hydrogen transport through two-dimensional monolayers
Seel, Max; Pandey, Ravindra
2016-06-01
Diffusion of protons and hydrogen atoms in representative two-dimensional materials is investigated. Specifically, density functional calculations were performed on graphene, hexagonal boron nitride (h-BN), phosphorene, silicene, and molybdenum disulfide (MoS2) monolayers to study the surface interaction and penetration barriers for protons and hydrogen atoms employing finite cluster models. The calculated barrier heights correlate approximately with the size of the opening formed by the three-fold open sites in the monolayers considered. They range from 1.56 eV (proton) and 4.61 eV (H) for graphene to 0.12 eV (proton) and 0.20 eV (H) for silicene. The results indicate that only graphene and h-BN monolayers have the potential for membranes with high selective permeability. The MoS2 monolayer behaves differently: protons and H atoms become trapped between the outer S layers in the Mo plane in a well with a depth of 1.56 eV (proton) and 1.5 eV (H atom), possibly explaining why no proton transport was detected, suggesting MoS2 as a hydrogen storage material instead. For graphene and h-BN, off-center proton penetration reduces the barrier to 1.38 eV for graphene and 0.11 eV for h-BN. Furthermore, Pt acting as a substrate was found to have a negligible effect on the barrier height. In defective graphene, the smallest barrier for proton diffusion (1.05 eV) is found for an oxygen-terminated defect. Therefore, it seems more likely that thermal protons can penetrate a monolayer of h-BN but not graphene and defects are necessary to facilitate the proton transport in graphene.
Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts
International Nuclear Information System (INIS)
Misra, B.; Maroni, V.A.
1978-01-01
A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated
Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts
International Nuclear Information System (INIS)
Misra, B.; Maroni, V.A.
1977-01-01
A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated
Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients
International Nuclear Information System (INIS)
Tuomisto, Harri.
1987-06-01
In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments
Moran, Robert P.
2013-01-01
Reactor fuel rod surface area that is perpendicular to coolant flow direction (+S) i.e. perpendicular to the P creates areas of coolant stagnation leading to increased coolant temperatures resulting in localized changes in fluid properties. Changes in coolant fluid properties caused by minor increases in temperature lead to localized reductions in coolant mass flow rates leading to localized thermal instabilities. Reductions in coolant mass flow rates result in further increases in local temperatures exacerbating changes to coolant fluid properties leading to localized thermal runaway. Unchecked localized thermal runaway leads to localized fuel melting. Reactor designs with randomized flow paths are vulnerable to localized thermal instabilities, localized thermal runaway, and localized fuel melting.
11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)
International Nuclear Information System (INIS)
Lemonnier, H.
2005-01-01
The main topics covered by the NURETH 11 meeting are the thermal-hydraulics of existing and future nuclear power plants as foreseen by the Generation IV worldwide initiative. Normal operation and accidental situations are also relevant topics of the Conference. The topics cover modeling, experiments, instrumentation and numerical simulations related to flow and heat transfer in nuclear reactors with a special emphasis on the advances of multiphase CFD methods. The first part of this Book of Abstracts enumerates the Organizing Scientific Societies, the Sponsors of the Conference, the Conference Chairs, and the members of the Steering Committee and of the Technical Program Committee. The second part of this Book of Abstracts contains the list of the titles of the contributed papers. Each item includes the log number of the paper, the abstract of which can therefore be easily located in the next section of this book. The titles of the papers have been sorted out by topics to provide a synthetic view of the contributions in a selected domain. The last section of this Book includes an index of authors and co-authors with a reference to the log number(s) of their contributed paper(s). Finally, the CD-Rom of the Conference Proceedings containing the full-length papers is inserted at the inside back cover. Sessions content: A - two-phase flow and heat transfer fundamentals: computational and mathematical techniques (numerical schemes, LBM, BEM, mesh-less, etc.); contact angle and wettability phenomena; experiments and data bases for the assessment and the verification of 3D models; flow regime identification and modelling; heat transfer near critical pressure and supercritical water reactors; interfacial area (data base, modeling, measurement techniques); instrumentation techniques; micro-scale basic phenomena, fluid flow and heat transfer; scaling methods; counter current flow; B - code developments: containment analysis; core thermal-hydraulics and subchannel analysis
EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques
Energy Technology Data Exchange (ETDEWEB)
Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1958-07-01
The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)
UK experience of safety requirements for thermal reactor stations
International Nuclear Information System (INIS)
Matthews, R.R.; Dale, G.C.; Tweedy, J.N.
1977-01-01
The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to
Review of the nuclear reactor thermal hydraulic research in ocean motions
International Nuclear Information System (INIS)
Yan, B.H.
2017-01-01
The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.
Review of the nuclear reactor thermal hydraulic research in ocean motions
Energy Technology Data Exchange (ETDEWEB)
Yan, B.H., E-mail: yanbh3@mail.sysu.edu.cn
2017-03-15
The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.
Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector
Energy Technology Data Exchange (ETDEWEB)
Bernstein, A; Bowden, N; Misner, A; Palmer, T
2007-06-27
In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7 days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.
The Choice of thermal reactor systems. A report by the National Nuclear Corporation Limited
International Nuclear Information System (INIS)
1977-01-01
The report to the Secretary of State in Great Britain by the National Nuclear Corporation following their assessment of the three thermal reactor systems, the AGR, PWR and SGHWR type reactors, which was performed in order to assist in the decision on the choice of thermal reactors for the U.K., is in three parts. Part I is an assessment of the three systems. It comprises: a description of the general method of assessment; a commentary in which are summarised discussions on the most important issues influencing reactor choice, i.e. safety, component failure, operational characteristics, development programme, construction programme; implications for the U.K. industry; costs; and reference design of each system. Part II consists of related questions and answers accompanied by commentaries on public acceptability and views from industry. Part III contains some conclusions including an analysis on the implications of the choices open and a summary of the main features of the assessment. (U.K.)
Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights
International Nuclear Information System (INIS)
Ninokata, H.; Kamide, H.
2011-01-01
In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)
Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor
Energy Technology Data Exchange (ETDEWEB)
Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A
2008-10-29
The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal
International Nuclear Information System (INIS)
Borovkov, A.I.; Semenov, A.S.; Granovsky, V.S.; Kovtunova, S.V.
1995-01-01
The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs
Energy Technology Data Exchange (ETDEWEB)
Borovkov, A.I.; Semenov, A.S. [St. Petersburg State Technical Univ. (Russian Federation); Granovsky, V.S.; Kovtunova, S.V. [Research Inst. of Technology, Sosnovy Bor (Russian Federation)
1995-12-31
The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs.
Experiments and analysis of thermal stresses around the nozzle of the reactor vessel
International Nuclear Information System (INIS)
Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.
1981-01-01
This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively
TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide
International Nuclear Information System (INIS)
Kaczynski, G.M.; Woodruff, R.W.
1985-12-01
TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2
Thermal-Hydraulic Analysis of a Supercritical Water Reactor (SCWR) Core
International Nuclear Information System (INIS)
Kucukboyaci, V.N.; Oriani, L.
2004-01-01
The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor
Monitoring device for the thermal margin of nuclear reactors
International Nuclear Information System (INIS)
Yoshikawa, Tatsuo
1984-01-01
Purpose: To extend the operation region and insure the stability thereby significantly improve the operation performance of a nuclear reactor by properly calculating a limited value for the minimum critical power ratio (OLMCPR) reflecting the actual reactor core state. Constitution: The device comprises a nuclear constant calculator, an abnormal transient analyzer and a transient critical power calculator. The abnormal transient analyzer performs analysis for the abnormal transient phenomena with a large variation amount of the minimum critical power ratio using the nuclear constants calculated by the nuclear constant calculator, to thereby determine transient changes such as the flow rate, power, pressure and entrance enthalpy of the reactor core. The transient critical power calculator determines the limited value for the minimum critical power ratio reflecting the state of the reactor core at the time to be monitored based on the thus determined transient change and display the same. Even if the value of MCPR determined by the process computer is smaller than the value for the designed OLMCPR, if it is greater than the displayed OLMCPR, procession such as power distribution control is unnecessary. (Nakamoto, H.)
Fast reactors. Thermal calculations of annulus application to Phenix
International Nuclear Information System (INIS)
Kung, J.P.; Gama, J.M.
1975-01-01
The gas convection phenomena involved in the annuli of the penetrations of the heat exchanger of the Phenix reactor are analyzed and the calculations performed using the BINIX program developed by GAAA to study the same phenomena are presented. The theory/experience comparison led to a better understanding of thermo-siphon phenomena [fr
Thermal performance of an insulating structure for a reactor vessel
International Nuclear Information System (INIS)
Aranovitch, E.; Crutzen, S.; LeDet, M.; Denis, R.
This report describes the installations used to test the HTGR reactor vessel insulating structure called ''Casali'' and details the experimental results in 3 groups: general experiments, systematic study, and technological experiments. The results obtained make it possible to satisfactorily predict the behavior of the structure in a practical application
Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor
International Nuclear Information System (INIS)
Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.
1994-05-01
The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan
N Reactor thermal plume characterization during Pu-only mode of operation
Energy Technology Data Exchange (ETDEWEB)
Ecker, R.M.; Thompson, F.L.; Whelan, G.
1983-04-01
Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.
Thermal hydraulic aspects of uncertainty in power measurement of nuclear reactors
International Nuclear Information System (INIS)
Gupta, S.K.; Kumar, Rajesh; Gaikwad, A.J.; Majumdar, P.; Agrawal, R.A.
2004-01-01
Power measurement in Nuclear Reactors is carried out through in-core and ex-core neutron monitors which are continuously calibrated against thermal power. In Indian Pressurized Heavy Water Reactors (220 MWe) the temperature difference across steam generator hot and cold legs is taken to be a measure of thermal power as the flow through the primary heat transport system is assumed to be constant through out is operation. Gross flow is not measured directly. However, the flow depends on the characteristics of the primary heat transport pumps, which are centrifugal type and are affected by the grid frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable high grid frequency. This uncertainty is in addition to instrument inaccuracy and should be accounted for in safety analysis. In some reactors thermal power is calculated from stem flow rate and pressure, here the location of steam flow measurement is important to avoid leakage related error in thermal power. Neutron absorption cross section in the power measurement instruments and the power production in the fuel varies with neutron energy levels, these aspects are also discussed in the paper. (author)
International Nuclear Information System (INIS)
Allen Hiser, J.R.
1993-01-01
In the framework of updating and clarification of the fracture toughness and thermal annealing requirements and guidance for light water reactor pressure vessels, proposed revisions concerning the pressurized thermal shock rule, fracture toughness requirements and reactor vessel material surveillance program requirements, are described. A new rule concerning thermal annealing requirements and a draft regulatory guide on 'Format and Content of Application for Approval for Thermal Annealing of RPV' are also proposed
Energy Technology Data Exchange (ETDEWEB)
Allen Hiser, J R [UKAEA Harwell Lab. (United Kingdom). Engineering Div.
1994-12-31
In the framework of updating and clarification of the fracture toughness and thermal annealing requirements and guidance for light water reactor pressure vessels, proposed revisions concerning the pressurized thermal shock rule, fracture toughness requirements and reactor vessel material surveillance program requirements, are described. A new rule concerning thermal annealing requirements and a draft regulatory guide on `Format and Content of Application for Approval for Thermal Annealing of RPV` are also proposed.
Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator
Garber, Anne E.; Dickens, Ricky E.
2011-01-01
The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.
Investigation on in-vessel thermal transients in a fast breeder reactor
International Nuclear Information System (INIS)
Muramatsu, Toshiharu; Kasahara, Naoto
1999-01-01
Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)
Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation
International Nuclear Information System (INIS)
Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia
2017-01-01
MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)
Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation
Energy Technology Data Exchange (ETDEWEB)
Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail: ghnfernandes@gmail.com, E-mail: marc5663@gmail.com, E-mail: athos1495@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)
2017-07-01
MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)
A contribution to the method of fast reactor thermal output calculation
International Nuclear Information System (INIS)
Harant, M.
1978-01-01
The method of stating the heat sources is discussed as being one of the factors influencing the accuracy of the thermal output calculation of fast reactors. The distribution of heat sources in the core and in other inner parts of the fast reactor is described using the least square fit method. Relations are derived of outputs of both individual components of fuel elements and of whole inner parts of the reactor. A comparison is made of various methods used for obtaining source integrals. The optimum integration method was found. (author)