A Methodology for Measuring Microplastic Transport in Large or Medium Rivers

Directory of Open Access Journals (Sweden)

Marcel Liedermann

2018-04-01

Full Text Available Plastic waste as a persistent contaminant of our environment is a matter of increasing concern due to the largely unknown long-term effects on biota. Although freshwater systems are known to be the transport paths of plastic debris to the ocean, most research has been focused on marine environments. In recent years, freshwater studies have advanced rapidly, but they rarely address the spatial distribution of plastic debris in the water column. A methodology for measuring microplastic transport at various depths that is applicable to medium and large rivers is needed. We present a new methodology offering the possibility of measuring microplastic transport at different depths of verticals that are distributed within a profile. The net-based device is robust and can be applied at high flow velocities and discharges. Nets with different sizes (41 µm, 250 µm, and 500 µm are exposed in three different depths of the water column. The methodology was tested in the Austrian Danube River, showing a high heterogeneity of microplastic concentrations within one cross section. Due to turbulent mixing, the different densities of the polymers, aggregation, and the growth of biofilms, plastic transport cannot be limited to the surface layer of a river, and must be examined within the whole water column as for suspended sediments. These results imply that multipoint measurements are required for obtaining the spatial distribution of plastic concentration and are therefore a prerequisite for calculating the passing transport. The analysis of filtration efficiency and side-by-side measurements with different mesh sizes showed that 500 µm nets led to optimal results.

Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard

2003-01-01

Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

Shielding calculations in support of the Spallation Neutron Source (SNS) proton beam transport system

International Nuclear Information System (INIS)

Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina

2002-01-01

Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system

Feasibility study on embedded transport core calculations

International Nuclear Information System (INIS)

Ivanov, B.; Zikatanov, L.; Ivanov, K.

2007-01-01

The main objective of this study is to develop an advanced core calculation methodology based on embedded diffusion and transport calculations. The scheme proposed in this work is based on embedded diffusion or SP 3 pin-by-pin local fuel assembly calculation within the framework of the Nodal Expansion Method (NEM) diffusion core calculation. The SP 3 method has gained popularity in the last 10 years as an advanced method for neutronics calculation. NEM is a multi-group nodal diffusion code developed, maintained and continuously improved at the Pennsylvania State University. The developed calculation scheme is a non-linear iteration process, which involves cross-section homogenization, on-line discontinuity factors generation, and boundary conditions evaluation by the global solution passed to the local calculation. In order to accomplish the local calculation, a new code has been developed based on the Finite Elements Method (FEM), which is capable of performing both diffusion and SP 3 calculations. The new code will be used in the framework of the NEM code in order to perform embedded pin-by-pin diffusion and SP 3 calculations on fuel assembly basis. The development of the diffusion and SP 3 FEM code is presented first following by its application to several problems. Description of the proposed embedded scheme is provided next as well as the obtained preliminary results of the C3 MOX benchmark. The results from the embedded calculations are compared with direct pin-by-pin whole core calculations in terms of accuracy and efficiency followed by conclusions made about the feasibility of the proposed embedded approach. (authors)

Range calculations using multigroup transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1979-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

International Nuclear Information System (INIS)

Carluccio, Thiago

2011-01-01

This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k eff and k src , and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

Development of a consistent Monte Carlo-deterministic transport methodology based on the method of characteristics and MCNP5

International Nuclear Information System (INIS)

Karriem, Z.; Ivanov, K.; Zamonsky, O.

2011-01-01

This paper presents work that has been performed to develop an integrated Monte Carlo- Deterministic transport methodology in which the two methods make use of exactly the same general geometry and multigroup nuclear data. The envisioned application of this methodology is in reactor lattice physics methods development and shielding calculations. The methodology will be based on the Method of Long Characteristics (MOC) and the Monte Carlo N-Particle Transport code MCNP5. Important initial developments pertaining to ray tracing and the development of an MOC flux solver for the proposed methodology are described. Results showing the viability of the methodology are presented for two 2-D general geometry transport problems. The essential developments presented is the use of MCNP as geometry construction and ray tracing tool for the MOC, verification of the ray tracing indexing scheme that was developed to represent the MCNP geometry in the MOC and the verification of the prototype 2-D MOC flux solver. (author)

Development and application of a hybrid transport methodology for active interrogation systems

Energy Technology Data Exchange (ETDEWEB)

Royston, K.; Walters, W.; Haghighat, A. [Nuclear Engineering Program, Department of Mechanical Engineering, Virginia Tech., 900 N Glebe Rd., Arlington, VA 22203 (United States); Yi, C.; Sjoden, G. [Nuclear and Radiological Engineering, Georgia Tech, 801 Ferst Drive, Atlanta, GA 30332 (United States)

2013-07-01

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)

Urban planning and industry in Spain: A novel methodology for calculating industrial carbon footprints

International Nuclear Information System (INIS)

Zubelzu, Sergio; Álvarez, Roberto

2015-01-01

In this paper we present a methodology for calculating the carbon footprint of the industrial sector during the urban planning stage in order to clearly develop and implement preventive measures. The methodology created focuses on industrial urban planning procedures and takes into account urban infrastructure in the characterization of GHG emissions. It allows for the implementation of preventive measures based on sustainability design criteria. The methodology was derived for specific industrial activity categories and was tested on a group of municipalities in a province south of Madrid, Spain. The results indicate that the average carbon footprint of industrial activities varies between 137.36 kgCO 2eq /m 2 e and 607.25 kgCO 2eq /m 2 e depending on the activity. Gas and electricity are the most important emissions sources for the most polluting industrial activities (chemical and nonmetal mineral products), while transportation is the most important source for every other activity. Municipalities can have a decisive influence on the industrial carbon footprint because, except for waste management and two industrial activities related to electricity, the majority of reductions can be achieved through urban planning decision variables. -- Highlights: •Model to calculate industrial carbon footprint in urban planning stage is proposed. •Specific industrial activities planned have a strong effect on carbon footprint. •Gas and electricity are the most relevant sources for the most pollutant industries. •Transport is relevant source for the less pollutant industries. •Municipalities can decisively influence on industrial carbon footprint

Preliminary integrated calculation of radionuclide cation and anion transport at Yucca Mountain using a geochemical model

International Nuclear Information System (INIS)

Birdsell, K.H.; Campbell, K.; Eggert, K.G.; Travis, B.J.

1989-01-01

This paper presents preliminary transport calculations for radionuclide movement at Yucca Mountain using preliminary data for mineral distributions, retardation parameter distributions, and hypothetical recharge scenarios. These calculations are not performance assessments, but are used to study the effectiveness of the geochemical barriers at the site at mechanistic level. The preliminary calculations presented have many shortcomings and should be viewed only as a demonstration of the modeling methodology. The simulations were run with TRACRN, a finite-difference porous flow and radionuclide transport code developed for the Yucca Mountain Project. Approximately 30,000 finite-difference nodes are used to represent the unsaturated and saturated zones underlying the repository in three dimensions. Sorption ratios for the radionuclides modeled are assumed to be functions of mineralogic assemblages of the underlying rock. These transport calculations present a representative radionuclide cation, 135 Cs and anion, 99 Tc. The effects on transport of many of the processes thought to be active at Yucca Mountain may be examined using this approach. The model provides a method for examining the integration of flow scenarios, transport, and retardation processes as currently understood for the site. It will also form the basis for estimates of the sensitivity of transport calculations to retardation processes. 11 refs., 17 figs., 1 tab

Audit calculation for the LOCA methodology for KSNP

Energy Technology Data Exchange (ETDEWEB)

Lee, Un Chul; Park, Chang Hwan; Choi, Yong Won; Yoo, Jun Soo [Seoul National Univ., Seoul (Korea, Republic of)

2006-11-15

The objective of this research is to perform the audit regulatory calculation for the LOCA methodology for KSNP. For LBLOCA calculation, several uncertainty variables and new ranges of those are added to those of previous KINS-REM to improve the applicability of KINS-REM for KSNP LOCA. And those results are applied to LBLOCA audit calculation by statistical method. For SBLOCA calculation, after selecting BATHSY9.1.b, which is not used by KHNP, the results of RELAP5/Mod3.3 and RELAP5/MOD3.3ef-sEM for KSNP SBLOCA are compared to evaluate the conservativeness or applicability of RELAP5/MOD3.3ef-sEM code for KSNP SBLOCA. The result of this research can be used to support the activities of KINS for reviewing the LOCA methodology for KSNP proposed by KHNP.

SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

International Nuclear Information System (INIS)

Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir

2014-01-01

The capabilities and limitations of SCALE6/MAVRIC hybrid deterministic-stochastic shielding methodology (CADIS and FW-CADIS) are demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by preparation of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR parameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in this paper was determination of neutron-gamma dose rate distribution (radiation field) over large portions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radiation included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical parameters of deterministic module plays important role for computer memory management. We investigated the possibility of using deterministic module (memory intense) with broad group library v7 2 7n19g opposed to fine group library v7 2 00n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with

Development of a database system for the calculation of indicators of environmental pressure caused by transport

DEFF Research Database (Denmark)

Giannouli, Myrsini; Samaras, Zissis; Keller, Mario

2006-01-01

The scope of this paper is to summarise a methodology developed for TRENDS (TRansport and ENvironment Database System-TRENDS). The main objective of TRENDS was the calculation of environmental pressure indicators caused by transport. The environmental pressures considered are associated with air...... emissions from the four main transport modes, i.e. road, rail, ships and air. In order to determine these indicators a system for calculating a range of environmental pressures due to transport was developed within a PC-based MS Access environment. Emphasis is given oil the latest features incorporated...... the production of collective results for all transport modes as well as a comparative assessment of air emissions produced by the various modes. Traffic activity and emission data obtained according to a basic (reference) scenario are displayed for the time period 1970-2020. In addition, a detailed assessment...

CALCULATION OF POLLUTION DYNAMICS NEAR RAILWAY TERRITORY DURING COAL TRANSPORTATION

Directory of Open Access Journals (Sweden)

M. M. Biliaiev

2017-02-01

Full Text Available Purpose. The article is aimed to develop 3D numerical model for the prediction of atmospheric pollution during transportation of bulk cargo in the railway car. Methodology.To solve this problem, it was developed three-dimensional numerical model, based on the use of the transport equation of dust pollution in the air by the wind and atmospheric turbulent diffusion. For the numerical integration of the simulating equation of the dust transport the implicit difference scheme was used. When constructing a difference scheme, it was carried out prior splitting of the original transport equation into the sequence of solutions of three equations. The first of them takes into account the transport of dust in paths, the second equation – dust transport under the influence of atmospheric turbulent diffusion, and the third equation –change of the dust concentration in the air due to its emissions from the cars.Unknown value of the pollutant concentration at every step of splitting is determined by the explicit scheme – the method of running account, which provides a simple numerical implementation of splitting equations. The developed numerical model is the basis for specialized computer program. On the basis of the constructed numerical model we carried out a computational experiment to assess the level of air pollution at the railway station during the motion of train with coal. Findings. Authors developed 3D numerical model, which belongs to the class of «screening models». This model takes into account the main physical factors affecting the process of dispersion of dust pollution in the atmosphere during coal transportation. The proposed numerical model requires low cost of computer time in the practical implementation on small and medium-power computers. This model can be used for rapid calculations of the dynamics of air pollution when transporting coal by rail. Calculations to determine the pollutant concentration and formation of the

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1977-01-01

A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount

Analysis of offsite dose calculation methodology for a nuclear power reactor

International Nuclear Information System (INIS)

Moser, D.M.

1995-01-01

This technical study reviews the methodology for calculating offsite dose estimates as described in the offsite dose calculation manual (ODCM) for Pennsylvania Power and Light - Susquehanna Steam Electric Station (SSES). An evaluation of the SSES ODCM dose assessment methodology indicates that it conforms with methodology accepted by the US Nuclear Regulatory Commission (NRC). Using 1993 SSES effluent data, dose estimates are calculated according to SSES ODCM methodology and compared to the dose estimates calculated according to SSES ODCM and the computer model used to produce the reported 1993 dose estimates. The 1993 SSES dose estimates are based on the axioms of Publication 2 of the International Commission of Radiological Protection (ICRP). SSES Dose estimates based on the axioms of ICRP Publication 26 and 30 reveal the total body estimates to be the most affected

Analysis of simulation methodology for calculation of the heat of transport for vacancy thermodiffusion

Energy Technology Data Exchange (ETDEWEB)

Tucker, William C.; Schelling, Patrick K., E-mail: patrick.schelling@ucf.edu [Advanced Material Processing and Analysis Center and Department of Physics, University of Central Florida, 4000 Central Florida Blvd., Orlando, Florida 32816 (United States)

2014-07-14

Computation of the heat of transport Q{sub a}{sup *} in monatomic crystalline solids is investigated using the methodology first developed by Gillan [J. Phys. C: Solid State Phys. 11, 4469 (1978)] and further developed by Grout and coworkers [Philos. Mag. Lett. 74, 217 (1996)], referred to as the Grout-Gillan method. In the case of pair potentials, the hopping of a vacancy results in a heat wave that persists for up to 10 ps, consistent with previous studies. This leads to generally positive values for Q{sub a}{sup *} which can be quite large and are strongly dependent on the specific details of the pair potential. By contrast, when the interactions are described using the embedded atom model, there is no evidence of a heat wave, and Q{sub a}{sup *} is found to be negative. This demonstrates that the dynamics of vacancy hopping depends strongly on the details of the empirical potential. However, the results obtained here are in strong disagreement with experiment. Arguments are presented which demonstrate that there is a fundamental error made in the Grout-Gillan method due to the fact that the ensemble of states only includes successful atom hops and hence does not represent an equilibrium ensemble. This places the interpretation of the quantity computed in the Grout-Gillan method as the heat of transport in doubt. It is demonstrated that trajectories which do not yield hopping events are nevertheless relevant to computation of the heat of transport Q{sub a}{sup *}.

Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

Energy Technology Data Exchange (ETDEWEB)

Richebois, E

2000-07-01

The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

Energy Technology Data Exchange (ETDEWEB)

Richebois, E

2000-07-01

The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

Relative Hazard Calculation Methodology

International Nuclear Information System (INIS)

DL Strenge; MK White; RD Stenner; WB Andrews

1999-01-01

The methodology presented in this document was developed to provide a means of calculating the RH ratios to use in developing useful graphic illustrations. The RH equation, as presented in this methodology, is primarily a collection of key factors relevant to understanding the hazards and risks associated with projected risk management activities. The RH equation has the potential for much broader application than generating risk profiles. For example, it can be used to compare one risk management activity with another, instead of just comparing it to a fixed baseline as was done for the risk profiles. If the appropriate source term data are available, it could be used in its non-ratio form to estimate absolute values of the associated hazards. These estimated values of hazard could then be examined to help understand which risk management activities are addressing the higher hazard conditions at a site. Graphics could be generated from these absolute hazard values to compare high-hazard conditions. If the RH equation is used in this manner, care must be taken to specifically define and qualify the estimated absolute hazard values (e.g., identify which factors were considered and which ones tended to drive the hazard estimation)

Methodologies of Uncertainty Propagation Calculation

International Nuclear Information System (INIS)

Chojnacki, Eric

2002-01-01

After recalling the theoretical principle and the practical difficulties of the methodologies of uncertainty propagation calculation, the author discussed how to propagate input uncertainties. He said there were two kinds of input uncertainty: - variability: uncertainty due to heterogeneity, - lack of knowledge: uncertainty due to ignorance. It was therefore necessary to use two different propagation methods. He demonstrated this in a simple example which he generalised, treating the variability uncertainty by the probability theory and the lack of knowledge uncertainty by the fuzzy theory. He cautioned, however, against the systematic use of probability theory which may lead to unjustifiable and illegitimate precise answers. Mr Chojnacki's conclusions were that the importance of distinguishing variability and lack of knowledge increased as the problem was getting more and more complex in terms of number of parameters or time steps, and that it was necessary to develop uncertainty propagation methodologies combining probability theory and fuzzy theory

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1976-01-01

A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount. For example, a deuterium plasma with 1.3 percent oxygen, one of the particle transport coefficients is increased by a factor of about four. The transport coefficients for the toroidal magnetic flux are reduced by about 20 percent. The increase in the particle transport coefficient is due to the collisional scattering of the deuterons by the heavy oxygen ions which is larger than the deuteron electron scattering, the normal process for particle transport in a two species plasma. The reduction in the toroidal magnetic flux transport coefficients are left unexplained

Development of Audit Calculation Methodology for RIA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Kim, Gwanyoung; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

The interim criteria contain more stringent limits than previous ones. For example, pellet-to-cladding mechanical interaction(PCMI) was introduced as a new failure criteria. And both short-term (e.g. fuel-to coolant interaction, rod burst) and long-term(e.g., fuel rod ballooning, flow blockage) phenomena should be addressed for core coolability assurance. For dose calculations, transient-induced fission gas release has to be accounted additionally. Traditionally, the approved RIA analysis methodologies for licensing application are developed based on conservative approach. But newly introduced safety criteria tend to reduce the margins to the criteria. Thereby, licensees are trying to improve the margins by utilizing a less conservative approach. In this situation, to cope with this trend, a new audit calculation methodology needs to be developed. In this paper, the new methodology, which is currently under developing in KINS, was introduced. For the development of audit calculation methodology of RIA safety analysis based on the realistic evaluation approach, preliminary calculation by utilizing the best estimate code has been done on the initial core of APR1400. Followings are main conclusions. - With the assumption of single full-strength control rod ejection in HZP condition, rod failure due to PCMI is not predicted. - And coolability can be assured in view of entalphy and fuel melting. - But, rod failure due to DNBR is expected, and there is possibility of fuel failure at the rated power conditions also.

Risk analysis methodologies for the transportation of radioactive materials

International Nuclear Information System (INIS)

Geffen, C.A.

1983-05-01

Different methodologies have evolved for consideration of each of the many steps required in performing a transportation risk analysis. Although there are techniques that attempt to consider the entire scope of the analysis in depth, most applications of risk assessment to the transportation of nuclear fuel cycle materials develop specific methodologies for only one or two parts of the analysis. The remaining steps are simplified for the analyst by narrowing the scope of the effort (such as evaluating risks for only one material, or a particular set of accident scenarios, or movement over a specific route); performing a qualitative rather than a quantitative analysis (probabilities may be simply ranked as high, medium or low, for instance); or assuming some generic, conservative conditions for potential release fractions and consequences. This paper presents a discussion of the history and present state-of-the-art of transportation risk analysis methodologies. Many reports in this area were reviewed as background for this presentation. The literature review, while not exhaustive, did result in a complete representation of the major methods used today in transportation risk analysis. These methodologies primarily include the use of severity categories based on historical accident data, the analysis of specifically assumed accident sequences for the transportation activity of interest, and the use of fault or event tree analysis. Although the focus of this work has generally been on potential impacts to public groups, some effort has been expended in the estimation of risks to occupational groups in transportation activities

Meta-Analytical Studies in Transport Economics. Methodology and Applications

Energy Technology Data Exchange (ETDEWEB)

Brons, M.R.E.

2006-05-18

Vast increases in the external costs of transport in the late twentieth century have caused national and international governmental bodies to worry about the sustainability of their transport systems. In this thesis we use meta-analysis as a research method to study various topics in transport economics that are relevant for sustainable transport policymaking. Meta-analysis is a research methodology that is based on the quantitative summarisation of a body of previously documented empirical evidence. In several fields of economic, meta-analysis has become a well-accepted research tool. Despite the appeal of the meta-analytical approach, there are methodological difficulties that need to be acknowledged. We study a specific methodological problem which is common in meta-analysis in economics, viz., within-study dependence caused by multiple sampling techniques. By means of Monte Carlo analysis we investigate the effect of such dependence on the performance of various multivariate estimators. In the applied part of the thesis we use and develop meta-analytical techniques to study the empirical variation in indicators of the price sensitivity of demand for aviation transport, the price sensitivity of demand for gasoline, the efficiency of urban public transport and the valuation of the external costs of noise from rail transport. We focus on the estimation of mean values for these indicators and on the identification of the impact of conditioning factors.

Acceleration methods for assembly-level transport calculations

International Nuclear Information System (INIS)

Adams, Marvin L.; Ramone, Gilles

1995-01-01

A family acceleration methods for the iterations that arise in assembly-level transport calculations is presented. A single iteration in these schemes consists of a transport sweep followed by a low-order calculation which is itself a simplified transport problem. It is shown that a previously-proposed method fitting this description is unstable in two and three dimensions. It is presented a family of methods and shown that some members are unconditionally stable. (author). 8 refs, 4 figs, 4 tabs

Methodology for calculating guideline concentrations for safety shot sites

International Nuclear Information System (INIS)

1997-06-01

Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination

Methodology for calculating guideline concentrations for safety shot sites

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination.

Electron stopping powers for transport calculations

International Nuclear Information System (INIS)

Berger, M.J.

1988-01-01

The reliability of radiation transport calculations depends on the accuracy of the input cross sections. Therefore, it is essential to review and update the cross sections from time to time. Even though the main interest of the author's group at NBS is in transport calculations and their applications, the group spends almost as much time on the analysis and preparation of cross sections as on the development of transport codes. Stopping powers, photon attenuation coefficients, bremsstrahlung cross sections, and elastic-scattering cross sections in recent years have claimed attention. This chapter deals with electron stopping powers (with emphasis on collision stopping powers), and reviews the state of the art as reflected by Report 37 of the International Commission on Radiation Units and Measurements

Three dimensions transport calculations for PWR core

International Nuclear Information System (INIS)

Richebois, E.

2000-01-01

The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

Radiation transport calculation methods in BNCT

International Nuclear Information System (INIS)

Koivunoro, H.; Seppaelae, T.; Savolainen, S.

2000-01-01

Boron neutron capture therapy (BNCT) is used as a radiotherapy for malignant brain tumours. Radiation dose distribution is necessary to determine individually for each patient. Radiation transport and dose distribution calculations in BNCT are more complicated than in conventional radiotherapy. Total dose in BNCT consists of several different dose components. The most important dose component for tumour control is therapeutic boron dose D B . The other dose components are gamma dose D g , incident fast neutron dose D f ast n and nitrogen dose D N . Total dose is a weighted sum of the dose components. Calculation of neutron and photon flux is a complex problem and requires numerical methods, i.e. deterministic or stochastic simulation methods. Deterministic methods are based on the numerical solution of Boltzmann transport equation. Such are discrete ordinates (SN) and spherical harmonics (PN) methods. The stochastic simulation method for calculation of radiation transport is known as Monte Carlo method. In the deterministic methods the spatial geometry is partitioned into mesh elements. In SN method angular integrals of the transport equation are replaced with weighted sums over a set of discrete angular directions. Flux is calculated iteratively for all these mesh elements and for each discrete direction. Discrete ordinates transport codes used in the dosimetric calculations are ANISN, DORT and TORT. In PN method a Legendre expansion for angular flux is used instead of discrete direction fluxes, land the angular dependency comes a property of vector function space itself. Thus, only spatial iterations are required for resulting equations. A novel radiation transport code based on PN method and tree-multigrid technique (TMG) has been developed at VTT (Technical Research Centre of Finland). Monte Carlo method solves the radiation transport by randomly selecting neutrons and photons from a prespecified boundary source and following the histories of selected particles

RAMA Methodology for the Calculation of Neutron Fluence

International Nuclear Information System (INIS)

Villescas, G.; Corchon, F.

2013-01-01

he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.

Methodology comparison for gamma-heating calculations in material-testing reactors

Energy Technology Data Exchange (ETDEWEB)

Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A. [CEA, DEN, DER, Cadarache F-13108 Saint Paul les Durance (France); Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France)

2015-07-01

The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear

A Hybrid Dynamic System Assessment Methodology for Multi-Modal Transportation-Electrification

Directory of Open Access Journals (Sweden)

Thomas J.T. van der Wardt

2017-05-01

Full Text Available In recent years, electrified transportation, be it in the form of buses, trains, or cars have become an emerging form of mobility. Electric vehicles (EVs, especially, are set to expand the amount of electric miles driven and energy consumed. Nevertheless, the question remains as to whether EVs will be technically feasible within infrastructure systems. Fundamentally, EVs interact with three interconnected systems: the (physical transportation system, the electric power grid, and their supporting information systems. Coupling of the two physical systems essentially forms a nexus, the transportation-electricity nexus (TEN. This paper presents a hybrid dynamic system assessment methodology for multi-modal transportation-electrification. At its core, it utilizes a mathematical model which consists of a marked Petri-net model superimposed on the continuous time microscopic traffic dynamics and the electrical state evolution. The methodology consists of four steps: (1 establish the TEN structure; (2 establish the TEN behavior; (3 establish the TEN Intelligent Transportation-Energy System (ITES decision-making; and (4 assess the TEN performance. In the presentation of the methodology, the Symmetrica test case is used throughout as an illustrative example. Consequently, values for several measures of performance are provided. This methodology is presented generically and may be used to assess the effects of transportation-electrification in any city or area; opening up possibilities for many future studies.

An optimized ultra-fine energy group structure for neutron transport calculations

International Nuclear Information System (INIS)

Huria, Harish; Ouisloumen, Mohamed

2008-01-01

This paper describes an optimized energy group structure that was developed for neutron transport calculations in lattices using the Westinghouse lattice physics code PARAGON. The currently used 70-energy group structure results in significant discrepancies when the predictions are compared with those from the continuous energy Monte Carlo methods. The main source of the differences is the approximations employed in the resonance self-shielding methodology. This, in turn, leads to ambiguous adjustments in the resonance range cross-sections. The main goal of developing this group structure was to bypass the self-shielding methodology altogether thereby reducing the neutronic calculation errors. The proposed optimized energy mesh has 6064 points with 5877 points spanning the resonance range. The group boundaries in the resonance range were selected so that the micro group cross-sections matched reasonably well with those derived from reaction tallies of MCNP for a number of resonance absorbers of interest in reactor lattices. At the same time, however, the fast and thermal energy range boundaries were also adjusted to match the MCNP reaction rates in the relevant ranges. The resulting multi-group library was used to obtain eigenvalues for a wide variety of reactor lattice numerical benchmarks and also the Doppler reactivity defect benchmarks to establish its adequacy. (authors)

Development of a simplified statistical methodology for nuclear fuel rod internal pressure calculation

International Nuclear Information System (INIS)

Kim, Kyu Tae; Kim, Oh Hwan

1999-01-01

A simplified statistical methodology is developed in order to both reduce over-conservatism of deterministic methodologies employed for PWR fuel rod internal pressure (RIP) calculation and simplify the complicated calculation procedure of the widely used statistical methodology which employs the response surface method and Monte Carlo simulation. The simplified statistical methodology employs the system moment method with a deterministic statistical methodology employs the system moment method with a deterministic approach in determining the maximum variance of RIP. The maximum RIP variance is determined with the square sum of each maximum value of a mean RIP value times a RIP sensitivity factor for all input variables considered. This approach makes this simplified statistical methodology much more efficient in the routine reload core design analysis since it eliminates the numerous calculations required for the power history-dependent RIP variance determination. This simplified statistical methodology is shown to be more conservative in generating RIP distribution than the widely used statistical methodology. Comparison of the significances of each input variable to RIP indicates that fission gas release model is the most significant input variable. (author). 11 refs., 6 figs., 2 tabs

Development of Monte Carlo decay gamma-ray transport calculation system

Energy Technology Data Exchange (ETDEWEB)

Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)

2001-06-01

In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)

The Methodology of Selecting the Transport Mode for Companies on the Slovak Transport Market

Science.gov (United States)

Černá, Lenka; Zitrický, Vladislav; Daniš, Jozef

2017-03-01

Transport volume in the Slovak Republic is growing continuously every year. This rising trend is influenced by the development of car industry and its suppliers. Slovak republic has also a geographic strategy position in middle Europe from the side of transport corridors (east-west and north-south). The development of transport volume in freight transport depends on the transport and business processes between the European Union and China and it is an opportunity for Slovak republic to obtain transit transport flows. In the Slovak Republic, road transport has a dominant position in the transport market. The volume of road transport has gradually increased over the past years. The increase of road transport is reflected on the highways and speed roads in regions which have higher economic potential. The increase of rail transport as seen on the main rail corridors is not as significant as in road transport. Trade globalization also has an influence on the increase of transport volume in intermodal transport. Predicted increase in transport volume for this transport mode is from 2,3 mil ton per year at present to 8 mil ton in the year 2020. Selection of transport mode and carrier is an important aspect for logistic management, because companies (customers) want to reduce the number of carriers which they trade and they create the system of several key carriers. Bigger transport volume and more qualitative transport service give a possibility to reduce transport costs. This trend is positive for carriers too, because the carriers can focus only on the selected customers and provide more qualitative services. The paper is focused on the selection of transport mode based on the proposed methodology. The aims of the paper are, definition of criteria which directly influence the selection of transport modes, determination of criteria based on the subjectively methods, creation of process for the selection of transport modes and practical application of proposed

CLEAR (Calculates Logical Evacuation And Response): A generic transportation network model for the calculation of evacuation time estimates

International Nuclear Information System (INIS)

Moeller, M.P.; Desrosiers, A.E.; Urbanik, T. II

1982-03-01

This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuation times for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies. (author)

CLEAR (Calculates Logical Evacuation And Response): A Generic Transportation Network Model for the Calculation of Evacuation Time Estimates

Energy Technology Data Exchange (ETDEWEB)

Moeller, M. P.; Urbanik, II, T.; Desrosiers, A. E.

1982-03-01

This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuatlon tlmes for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies.

New model for mines and transportation tunnels external dose calculation using Monte Carlo simulation

International Nuclear Information System (INIS)

Allam, Kh. A.

2017-01-01

In this work, a new methodology is developed based on Monte Carlo simulation for tunnels and mines external dose calculation. Tunnels external dose evaluation model of a cylindrical shape of finite thickness with an entrance and with or without exit. A photon transportation model was applied for exposure dose calculations. A new software based on Monte Carlo solution was designed and programmed using Delphi programming language. The variation of external dose due to radioactive nuclei in a mine tunnel and the corresponding experimental data lies in the range 7.3 19.9%. The variation of specific external dose rate with position in, tunnel building material density and composition were studied. The given new model has more flexible for real external dose in any cylindrical tunnel structure calculations. (authors)

Statistics of Monte Carlo methods used in radiation transport calculation

International Nuclear Information System (INIS)

Datta, D.

2009-01-01

Radiation transport calculation can be carried out by using either deterministic or statistical methods. Radiation transport calculation based on statistical methods is basic theme of the Monte Carlo methods. The aim of this lecture is to describe the fundamental statistics required to build the foundations of Monte Carlo technique for radiation transport calculation. Lecture note is organized in the following way. Section (1) will describe the introduction of Basic Monte Carlo and its classification towards the respective field. Section (2) will describe the random sampling methods, a key component of Monte Carlo radiation transport calculation, Section (3) will provide the statistical uncertainty of Monte Carlo estimates, Section (4) will describe in brief the importance of variance reduction techniques while sampling particles such as photon, or neutron in the process of radiation transport

A methodology for evaluating environmental impacts of railway freight transportation policies

International Nuclear Information System (INIS)

Lopez, Ignacio; Rodriguez, Javier; Buron, Jose Manuel; Garcia, Alberto

2009-01-01

Railway freight transportation presents a degree of complexity which frequently makes impossible to model it with sufficient precision. Currently, energetic and environmental impacts of freight transportation are usually modelled following average data, which do not reflect the characteristics of specific lines. These models allow qualitative approximations which may be used as criteria for designing high-level transportation policies: road-train modal shift, regional energetic planning or environmental policies. This paper proposes a methodology for estimating railway consumption associated to a specific railway line which yields a new degree of precision. It is based on estimating different contributions to railway consumption by a collection of factors, mobility, operation, or infrastructure-related. This procedure also allows applying the methodology for designing transportation policies in detail: evaluating impact of modal shift, consumption and pollutant emissions on a specific line, as well as the effect of building tunnels, reducing slopes, improving traffic control, etc. A comparison of the estimations given by the conventional approach and the proposed methodology is offered, as well as further comments on the results.

Calculation of external exposure during transport and disposal of radioactive waste arisen from dismantling of steam generator

International Nuclear Information System (INIS)

Hornacek, M.; Necas, V.

2014-01-01

The dismantling of large components (reactor pressure vessel, reactor internals, steam generator) represents complex of processes involving preparation, dismantling, waste treatment and conditioning, transport and final disposal. To optimise all of these activities in accordance with the ALARA principle the prediction of the exposure of workers is an essential prerequisite. The paper deals with the calculation of external exposure of workers during transport and final disposal of heat exchange tubes of steam generator used in Slovak nuclear power plant V1 in Jaslovske Bohunice. The type of waste packages, the calculation models of truck and National Radioactive Waste Repository in Mochovce are presented. The detailed methodology of radioactive waste disposal is showed and the degree of influence of time decay (0, 5 and 10 years) on the radiological conditions during transport and disposal is studied. All of the results do not exceed the limits given in Slovak and international regulatory documents. (authors)

Sn transport calculations on vector and parallel processors

International Nuclear Information System (INIS)

Rhoades, W.A.; Childs, R.L.

1987-01-01

The transport of radiation from the source to the location of people or equipment gives rise to some of the most challenging of calculations. A problem may involve as many as a billion unknowns, each evaluated several times to resolve interdependence. Such calculations run many hours on a Cray computer, and a typical study involves many such calculations. This paper will discuss the steps taken to vectorize the DOT code, which solves transport problems in two space dimensions (2-D); the extension of this code to 3-D; and the plans for extension to parallel processors

Methodology and conclusions of activation calculations of WWER-440 type nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Babcsány, Boglárka, E-mail: boglarka.babcsany@reak.bme.hu; Czifrus, Szabolcs; Fehér, Sándor

2015-04-01

Highlights: • Activation calculation of two WWER-440 type nuclear power plants. • Detailed description of the applied activation calculation methodology. • Graphical results for total activity and waste index categorization. • General conclusions for activation applicable in the case of PWR reactors. - Abstract: Activation calculations for two nuclear power plants of WWER-440 type have been performed by the authors in order to assist the decommissioning planning by assessing the radioactive inventory present at the time of and at different times after the final shutdown. According to related international literature and studies performed earlier by the authors, considering the activity more than 99% of this inventory is concentrated in the materials directly surrounding the reactor core, where the predominant evolution of radionuclides is generated by neutron induced nuclear reactions. In order to obtain the highest possible accuracy in modelling, three-dimensional Monte Carlo neutron transport calculations were performed. Besides the methods and models applied to these analyses, the paper also summarizes the results that can be generally applied to such nuclear power plant types. At the time of shutdown, the total activity of the stainless steel components is about 6 × 10{sup 16} Bq and 1.3 × 10{sup 17} Bq for the two NPPs considered. The biological shielding concrete constitutes approximately 7 × 10{sup 13} Bq and 1.1 × 10{sup 14} Bq.

Discrete-ordinates electron transport calculations using standard neutron transport codes

International Nuclear Information System (INIS)

Morel, J.E.

1979-01-01

The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

Increasing the competitiveness of maintenance contract rates by using an alternative methodology for the calculation of average vehicle maintenance costs

Directory of Open Access Journals (Sweden)

Stephen Carstens

2008-11-01

Full Text Available Companies tend to outsource transport to fleet management companies to increase efficiencies if transport is a non-core activity. The provision of fleet management services on contract introduces a certain amount of financial risk to the fleet management company, specifically fixed rate maintenance contracts. The quoted rate needs to be sufficient and also competitive in the market. Currently the quoted maintenance rates are based on the maintenance specifications of the manufacturer and the risk management approach of the fleet management company. This is usually reflected in a contingency that is included in the quoted maintenance rate. An alternative methodology for calculating the average maintenance cost for a vehicle fleet is proposed based on the actual maintenance expenditures of the vehicles and accepted statistical techniques. The proposed methodology results in accurate estimates (and associated confidence limits of the true average maintenance cost and can beused as a basis for the maintenance quote.

Methodology for calculating power consumption of planetary mixers

Science.gov (United States)

Antsiferov, S. I.; Voronov, V. P.; Evtushenko, E. I.; Yakovlev, E. A.

2018-03-01

The paper presents the methodology and equations for calculating the power consumption necessary to overcome the resistance of a dry mixture caused by the movement of cylindrical rods in the body of a planetary mixer, as well as the calculation of the power consumed by idling mixers of this type. The equations take into account the size and physico-mechanical properties of mixing material, the size and shape of the mixer's working elements and the kinematics of its movement. The dependence of the power consumption on the angle of rotation in the plane perpendicular to the axis of rotation of the working member is presented.

Alternative methodology for irradiation reactor experimental shielding calculation

International Nuclear Information System (INIS)

Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

1996-01-01

Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

Charged-particle calculations using Boltzmann transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Dodds, H.L. Jr.; Robinson, M.T.; Holmes, D.K.

1981-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of charged particle range distributions, reflection coefficients, and sputtering yields. The Boltzmann transport approach can be implemented, with minor changes, in standard neutral particle computer codes. With the multigroup discrete ordinates code, ANISN, determination of ion and target atom distributions as functions of position, energy, and direction can be obtained without the stochastic error associated with atomistic computer codes such as MARLOWE and TRIM. With the multigroup Monte Carlo code, MORSE, charged particle effects can be obtained for problems associated with very complex geometries. Results are presented for several charged particle problems. Good agreement is obtained between quantities calculated with the multigroup approach and those obtained experimentally or by atomistic computer codes

ExternE transport methodology for external cost evaluation of air pollution

DEFF Research Database (Denmark)

Jensen, S. S.; Berkowicz, R.; Brandt, J.

The report describes how the human exposure estimates based on NERI's human exposure modelling system (AirGIS) can improve the Danish data used for exposure factors in the ExternE Transport methodology. Initially, a brief description of the ExternE Tranport methodology is given and it is summarised...

User needs for a standardized CO2 emission assessment methodology for intelligent transport systems

NARCIS (Netherlands)

Mans, D.; Rekiel, J.; Wolfermann, A.; Klunder, G.

2012-01-01

The Amitran FP7 project will define a reference methodology to assess the impact of intelligent transport systems on CO2 emissions. The methodology is intended to be used as a reference by future projects and covers both passenger and freight transport. The project will lead to a validated

A methodology for calculating photovoltaic field output and effect of solar tracking strategy

International Nuclear Information System (INIS)

Hu, Yeguang; Yao, Yingxue

2016-01-01

Highlights: • A new methodology for calculating PV field output is proposed. • The reduction of diffuse radiation and albedo due to shading is considered. • The shadow behavior is accurately analyzed at a cell level. • Several simplified measures are taken to reduce the calculation work. • The field outputs with different solar tracking strategies are compared. - Abstract: This paper proposes an effective methodology for calculating the photovoltaic field output. A combination of two methods is first presented for optical performance calculation: point projection method for direction radiation, and Monte Carlo ray-tracing method for both diffuse radiation and albedo radiation. Based on the optical calculation, an accurate output of the photovoltaic field can be obtained through a cell-level simulation of PV system. Several simplified measures are taken to reduce the large amount of calculation work. The proposed methodology has been validated for accurate and fast calculation of field output. With the help of the developed code, this paper deals with the performance comparison between four typical tracking strategies. Through the comparative analysis, the field output is proved to be related to the tracking strategy. For a regular photovoltaic field, the equatorial and elevation-rolling tracking show the superior performance in annual field output to the azimuth-elevation and rolling-elevation tracking. A reasonable explanation for this difference has been presented in this paper.

An Application of the Methodology for Assessment of the Sustainability of Air Transport System

Science.gov (United States)

Janic, Milan

2003-01-01

An assessment and operationalization of the concept of sustainable air transport system is recognized as an important but complex research, operational and policy task. In the scope of the academic efforts to properly address the problem, this paper aims to assess the sustainability of air transport system. It particular, the paper describes the methodology for assessment of sustainability and its potential application. The methodology consists of the indicator systems, which relate to the air transport system operational, economic, social and environmental dimension of performance. The particular indicator systems are relevant for the particular actors such users (air travellers), air transport operators, aerospace manufacturers, local communities, governmental authorities at different levels (local, national, international), international air transport associations, pressure groups and public. In the scope of application of the methodology, the specific cases are selected to estimate the particular indicators, and thus to assess the system sustainability under given conditions.

Development of a database system for the calculation of indicators of environmental pressure caused by transport

Energy Technology Data Exchange (ETDEWEB)

Giannouli, Myrsini; Samaras, Zissis [Aristotle University of Thessaloniki, Laboratory of Applied Thermodynamics, Mechanical Engineering Department, GR 54124, Thessaloniki, P.O. Box 458 (Greece); Keller, Mario; De Haan, Peter [INFRAS, Muhlemattstrasse 45 CH-3007, Bern (Switzerland); Kallivoda, Manfred [psiA-Consult, Environmental Research and Engineering GmbH, Lastenstrasse 38/1, 1230 Wien (Austria); Sorenson, Spencer; Georgakaki, Aliki [DTU: Technical University of Denmark, Nils Koppels Alle, Building 403, DK 2800 Kgs. Lyngby (Denmark)

2006-03-15

The scope of this paper is to summarise a methodology developed for TRENDS (TRansport and ENvironment Database System-TRENDS). The main objective of TRENDS was the calculation of environmental pressure indicators caused by transport. The environmental pressures considered are associated with air emissions from the four main transport modes, i.e. road, rail, ships and air. In order to determine these indicators a system for calculating a range of environmental pressures due to transport was developed within a PC-based MS Access environment. Emphasis is given on the latest features incorporated in the model and their applications. One of the recently developed features of the software provides an option for simple scenario analysis including vehicle dynamics (such as turnover and evolution) for all EU15 member states. This feature is called the Transport Activity Balance module (TAB) and enables the production of collective results for all transport modes as well as a comparative assessment of air emissions produced by the various modes. Traffic activity and emission data obtained according to a basic (reference) scenario are displayed for the time period 1970-2020. In addition, a detailed assessment of the results produced by TRENDS was conducted by means of comparison with data found in the literature. Finally, vehicle emissions produced by the model for the EU15 member states were spatially disaggregated for the base year, 1995 and GIS maps were generated. Examples of these maps are displayed in this document, for the various modes of transport considered in the study. (author)

Conceptual and methodological approaches to evaluation of investment attractiveness of enterprises engaged in transportations

Directory of Open Access Journals (Sweden)

Olha Myshkovych

2016-12-01

Full Text Available The aim of the article is to analyze the conceptual and methodological approaches to determining the investment attractiveness of enterprises engaged in transportations. It is indicated that the investment attractiveness of transport enterprises should be determined by calculating of the overall financial situation of enterprises, which will allow potential investors to evaluate profitability and cost efficiency of its activity. An analysis of the strengths and weaknesses of the enterprise engaged in transportation can be accomplished by the evaluation of its innovative capacity. The identification of factors and reserves of the increasing of enterprise innovative development will allow distinguishing of the basic directions for the improvement of organizational and economic mechanism of its activity. With the aim of building the strategy for the strengthening of market position it is also considered important for the potential investor to obtain the information about enterprise place on the national and international markets. Political and legal environment, characterized by political stability of society and the regulatory framework of entrepreneurial and investment activity serve as a certain guarantee of the investment reliability.

Generalized diffusion theory for calculating the neutron transport scalar flux

International Nuclear Information System (INIS)

Alcouffe, R.E.

1975-01-01

A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)

Calculation of three-dimensional groundwater transport using second-order moments

International Nuclear Information System (INIS)

Pepper, D.W.; Stephenson, D.E.

1987-01-01

Groundwater transport of contaminants from the F-Area seepage basin at the Savannah River Plant (SRP) was calculated using a three-dimensional, second-order moment technique. The numerical method calculates the zero, first, and second moment distributions of concentration within a cell volume. By summing the moments over the entire solution domain, and using a Lagrangian advection scheme, concentrations are transported without numerical dispersion errors. Velocities obtained from field tests are extrapolated and interpolated to all nodal points; a variational analysis is performed over the three-dimensional velocity field to ensure mass consistency. Transport predictions are calculated out to 12,000 days. 28 refs., 9 figs

Molecular transport calculations with Wannier Functions

DEFF Research Database (Denmark)

Thygesen, Kristian Sommer; Jacobsen, Karsten Wedel

2005-01-01

We present a scheme for calculating coherent electron transport in atomic-scale contacts. The method combines a formally exact Green's function formalism with a mean-field description of the electronic structure based on the Kohn-Sham scheme of density functional theory. We use an accurate plane...

Methodology for Calculating Latency of GPS Probe Data

Energy Technology Data Exchange (ETDEWEB)

Young, Stanley E [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Wang, Zhongxiang [University of Maryland; Hamedi, Masoud [University of Maryland

2017-10-01

Crowdsourced GPS probe data, such as travel time on changeable-message signs and incident detection, have been gaining popularity in recent years as a source for real-time traffic information to driver operations and transportation systems management and operations. Efforts have been made to evaluate the quality of such data from different perspectives. Although such crowdsourced data are already in widespread use in many states, particularly the high traffic areas on the Eastern seaboard, concerns about latency - the time between traffic being perturbed as a result of an incident and reflection of the disturbance in the outsourced data feed - have escalated in importance. Latency is critical for the accuracy of real-time operations, emergency response, and traveler information systems. This paper offers a methodology for measuring probe data latency regarding a selected reference source. Although Bluetooth reidentification data are used as the reference source, the methodology can be applied to any other ground truth data source of choice. The core of the methodology is an algorithm for maximum pattern matching that works with three fitness objectives. To test the methodology, sample field reference data were collected on multiple freeway segments for a 2-week period by using portable Bluetooth sensors as ground truth. Equivalent GPS probe data were obtained from a private vendor, and their latency was evaluated. Latency at different times of the day, impact of road segmentation scheme on latency, and sensitivity of the latency to both speed-slowdown and recovery-from-slowdown episodes are also discussed.

ANL calculational methodologies for determining spent nuclear fuel source term

International Nuclear Information System (INIS)

McKnight, R. D.

2000-01-01

Over the last decade Argonne National Laboratory has developed reactor depletion methods and models to determine radionuclide inventories of irradiated EBR-II fuels. Predicted masses based on these calculational methodologies have been validated using available data from destructive measurements--first from measurements of lead EBR-II experimental test assemblies and later using data obtained from processing irradiated EBR-II fuel assemblies in the Fuel Conditioning Facility. Details of these generic methodologies are described herein. Validation results demonstrate these methods meet the FCF operations and material control and accountancy requirements

Hot channel calculation methodologies in case of Gd burnable poison

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2008-01-01

The final step in the safety analysis is the investigation of the fulfilment of the acceptance criteria using hot channel calculations. Recently, there has been under way at Paks NPP to introduce a new, higher enriched (4.2 %) fuel type containing Gd burnable poison. To do that, for some transients the DBA analyses must be repeated and last year, as one of the first steps in this process, it was needed to review the hot channel calculation methodologies used in the analyses. The goal of the paper is to summarize some aspects of the hot channel calculation methodologies using different lattice pitches and different fuel types (Gd or non Gd and different enrichments). Mainly, three topics are discussed. First, the influence of the radial power distribution (and other burnup dependent parameters) inside the fuel pin are investigated, and then we discuss the problem of the selection of the appropriate 'frame parameter' in connection with the initial power level at the initial stationary state of DBA transients. Finally, we are trying to answer the question: is it possible to build up a conservative single closed sub-channel approach against multi channel approach?(Authors)

Current evaluation of dose rate calculation - analytical method

International Nuclear Information System (INIS)

Tello, Marcos; Vilhena, Marco Tulio

1996-01-01

The accuracy of the dose calculations based on pencil beam formulas such as Fokker-Plank equations and Fermi equations for charged particle transport are studied and a methodology to solve the Boltzmann transport equation is suggested

76 FR 34270 - Federal-State Extended Benefits Program-Methodology for Calculating “on” or “off” Total...

Science.gov (United States)

2011-06-13

... requirement. The Department plans to promulgate regulations about this methodology in the near future. In the...--Methodology for Calculating ``on'' or ``off'' Total Unemployment Rate Indicators for Purposes of Determining..., Labor. ACTION: Notice. SUMMARY: UIPL 16-11 informs states of the methodology used to calculate the ``on...

Parallel SN transport calculations on a transputer network

International Nuclear Information System (INIS)

Kim, Yong Hee; Cho, Nam Zin

1994-01-01

A parallel computing algorithm for the neutron transport problems has been implemented on a transputer network and two reactor benchmark problems (a fixed-source problem and an eigenvalue problem) are solved. We have shown that the parallel calculations provided significant reduction in execution time over the sequential calculations

Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

International Nuclear Information System (INIS)

Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

2013-01-01

The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

Neutron transport assembly calculation with non-zero net current boundary condition

International Nuclear Information System (INIS)

Jo, Chang Keun

1993-02-01

Fuel assembly calculation for the homogenized group constants is one of the most important parts in the reactor core analysis. The homogenized group constants of one a quarter assembly are usually generated for the nodal calculation of the reactor core. In the current nodal calculation, one or a quarter of the fuel assembly corresponds to a unit node. The homogenized group constant calculation for a fuel assembly proceeds through cell spectrum calculations, group condensation and cell homogenization calculations, two dimensional fuel assembly calculation, and then depletion calculations of fuel rods. To obtain the assembly wise homogenized group constants, the two dimensional transport calculation is usually performed. Most codes for the assembly wise homogenized group constants employ a zero net current boundary condition. CASMO-3 is such a code that is in wide use. The zero net current boundary condition is plausible and valid in an infinite reactor composed of the same kind of assemblies. However, the reactor is finite and the core is constructed by different kinds of assemblies. Hence, the assumption of the zero net current boundary condition is not valid in the actual reactor. The objective of this study is to develop a homogenization methodology that can treat any actual boundary condition, i.e. non-zero net current boundary condition. In order to treat the non-zero net current boundary condition, we modify CASMO-3. For the two-dimensional treatment in CASMO-3, a multigroup integral transport routine based on the method of transmission probability is used. The code performs assembly calculation with zero net current boundary condition. CASMO-3 is modified to consider the inhomogeneous source at the assembly boundary surface due to the non-zero net current. The modified version of CASMO-3 is called CASMO-3M. CASMO-3M is applied to several benchmark problems. In order to obtain the inhomogeneous source, the global calculation is performed. The local calculation

ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

International Nuclear Information System (INIS)

1993-01-01

1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

Methodology of dose calculation for the SRS SAR

International Nuclear Information System (INIS)

Price, J.B.

1991-07-01

The Savannah River Site (SRS) Safety Analysis Report (SAR) covering K reactor operation assesses a spectrum of design basis accidents. The assessment includes estimation of the dose consequences from the analyzed accidents. This report discusses the methodology used to perform the dose analysis reported in the SAR and also includes the quantified doses. Doses resulting from postulated design basis reactor accidents in Chapter 15 of the SAR are discussed, as well as an accident in which three percent of the fuel melts. Doses are reported for both atmospheric and aqueous releases. The methodology used to calculate doses from these accidents as reported in the SAR is consistent with NRC guidelines and industry standards. The doses from the design basis accidents for the SRS reactors are below the limits set for commercial reactors by the NRC and also meet industry criteria. A summary of doses for various postulated accidents is provided

Proposal of risk evaluation methodology for hazardous materials transportation

International Nuclear Information System (INIS)

Hartman, Luiz Carlos

2009-01-01

The increasing concern with the level of risk associated with the transportation of hazardous materials took some international institutions to pledge efforts in the evaluation of risk in regional level. Following this trend, the objective of this work was to analyze the most recent processes of analysis of risks from road transportation of hazardous materials. In the present work 21 methodologies of analysis of risks, developed by some authors and for diverse localities have been evaluated. Two of them, in special, have been reviewed and discussed: a method recently developed by the Swiss Federal Institute of Technology (Nicolet-Monnier and Gheorghe, 1996) and the strategy delineated by the Center for Chemical Process Safety CCPS (1995), taking into consideration the estimate of the individual and social risk. Also, the models of Harwood et al. (1990) and of Ramos (1997), adapted by Hartman (2003) have been applied to the reality of the roads of the state of Sao Paulo. The extension of these methodologies was explored, in order to find its advantages and disadvantages. As a study case the present work considered the ammonia transportation throughout two routes evaluating the reality of the roads of the state of Sao Paulo, including a significant parcel of evaluation in a densely populated area, getting the results using risk, at least, one of the methodologies mentioned above. The innovation proposed by this work was the research, the development and the introduction of two variables to the model considered by Harwood et al. (1990). These variables that influence in the value of the risk are: the age of the driver of truck and the zone of impact that is function type of product, period of the day where the transport was carried and the volume that has been transported. The aim of the proposed modifications is to let the value of the risk more sensible in relation to the type of the product carried and the age of the truck driver. The main related procedural stages

42 CFR 484.230 - Methodology used for the calculation of the low-utilization payment adjustment.

Science.gov (United States)

2010-10-01

... 42 Public Health 5 2010-10-01 2010-10-01 false Methodology used for the calculation of the low... Prospective Payment System for Home Health Agencies § 484.230 Methodology used for the calculation of the low... amount is determined by using cost data set forth in § 484.210(a) and adjusting by the appropriate wage...

Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

Energy Technology Data Exchange (ETDEWEB)

Bordy, J M; Kodeli, I; Menard, St; Bouchet, J L; Renard, F; Martin, E; Blazy, L; Voros, S; Bochud, F; Laedermann, J P; Beaugelin, K; Makovicka, L; Quiot, A; Vermeersch, F; Roche, H; Perrin, M C; Laye, F; Bardies, M; Struelens, L; Vanhavere, F; Gschwind, R; Fernandez, F; Quesne, B; Fritsch, P; Lamart, St; Crovisier, Ph; Leservot, A; Antoni, R; Huet, Ch; Thiam, Ch; Donadille, L; Monfort, M; Diop, Ch; Ricard, M

2006-07-01

The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

Development of 3D pseudo pin-by-pin calculation methodology in ANC

International Nuclear Information System (INIS)

Zhang, B.; Mayhue, L.; Huria, H.; Ivanov, B.

2012-01-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000 R plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. The mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)

Minaret, a deterministic neutron transport solver for nuclear core calculations

Energy Technology Data Exchange (ETDEWEB)

Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

2011-07-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Minaret, a deterministic neutron transport solver for nuclear core calculations

International Nuclear Information System (INIS)

Moller, J-Y.; Lautard, J-J.

2011-01-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Regulatory guides for qualifying the calculation methodology of Furnas by CNEN

International Nuclear Information System (INIS)

1987-10-01

Regulatory guides are presented which will be used for qualifying the calculation methodology of FURNAS by CNEN, in the areas of Neutronics, Thermohydraulics, Accident Analysis and Fuel Rod Performance, as applied to Angra 1 NPP. (Author) [pt

A methodology for the evaluation of fuel rod failures under transportation accidents

International Nuclear Information System (INIS)

Rashid, J.Y.R.; Machiels, A.J.

2004-01-01

, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite element code. The characterization of hydrides-dependent properties of high-burnup fuel cladding is the main fea-ture of this constitutive model. The third element in the overall process is to utilize this material model and its host finite element code in the structural analysis of a transportation cask subjected to bounding accident loading to cal-culate fuel rod failures and failure mode configurations. This requires detailed modeling of the transport cask and its internal structure, which include canister, basket, fuel assembly grids and fuel rods. The overall methodology is described in the paper

Application of a numerical transport correction in diffusion calculations

International Nuclear Information System (INIS)

Tomatis, Daniele; Dall'Osso, Aldo

2011-01-01

Full core calculations by ordinary transport methods can demand considerable computational time, hardly acceptable in the industrial work frame. However, the trend of next generation nuclear cores goes toward more heterogeneous systems, where transport phenomena of neutrons become very important. On the other hand, using diffusion solvers is more practical allowing faster calculations, but a specific formulation of the diffusion coefficient is requested to reproduce the scalar flux with reliable physical accuracy. In this paper, the Ronen method is used to evaluate numerically the diffusion coefficient in the slab reactor. The new diffusion solution is driven toward the solution of the integral neutron transport equation by non linear iterations. Better estimates of currents are computed and diffusion coefficients are corrected at node interfaces, still assuming Fick's law. This method enables obtaining closer results to the transport solution by a common solver in multigroup diffusion. (author)

Development of the processing software package for RPV neutron fluence determination methodology

International Nuclear Information System (INIS)

Belousov, S.; Kirilova, K.; Ilieva, K.

2001-01-01

According to the INRNE methodology the neutron transport calculation is carried out by two steps. At the first step reactor core eigenvalue calculation is performed. This calculation is used for determination of the fixed source for the next step calculation of neutron transport from the reactor core to the RPV. Both calculation steps are performed by state of the art and tested codes. The interface software package DOSRC developed at INRNE is used as a link between these two calculations. The package transforms reactor core calculation results to neutron source input data in format appropriate for the neutron transport codes (DORT, TORT and ASYNT) based on the discrete ordinates method. These codes are applied for calculation of the RPV neutron flux and its responses - induced activity, radiation damage, neutron fluence etc. Fore more precise estimation of the neutron fluence, the INRNE methodology has been supplemented by the next improvements: - implementation of more advanced codes (PYTHIA/DERAB) for neutron-physics parameter calculations; - more detailed neutron source presentation; - verification of neutron fluence by statistically treated experimental data. (author)

Recent progress and developments in LWR-PV calculational methodology

International Nuclear Information System (INIS)

Maerker, R.E.; Broadhead, B.L.; Williams, M.L.

1984-01-01

New and improved techniques for calculating beltline surveillance activities and pressure vessel fluences with reduced uncertainties have recently been developed. These techniques involve the combining of monitored in-core power data with diffusion theory calculated pin-by-pin data to yield absolute source distributions in R-THETA and R-Z geometries suitable for discrete ordinate transport calculations. Effects of finite core height, whenever necessary, can be considered by the use of a three-dimensional fluence rate synthesis procedure. The effects of a time-dependent spatial source distribution may be readily evaluated by applying the concept of the adjoint function, and simplifying the procedure to such a degree that only one forward and one adjoint calculation are required to yield all the dosimeter activities for all beltline surveillance locations at once. The addition of several more adjoint calculations using various fluence rates as responses is all that is needed to determine all the pressure vessel group fluences for all beltline locations for an arbitrary source distribution

Efficient calculation of dissipative quantum transport properties in semiconductor nanostructures

Energy Technology Data Exchange (ETDEWEB)

Greck, Peter

2012-11-26

We present a novel quantum transport method that follows the non-equilibrium Green's function (NEGF) framework but side steps any self-consistent calculation of lesser self-energies by replacing them by a quasi-equilibrium expression. We termed this method the multi-scattering Buettiker-Probe (MSB) method. It generalizes the so-called Buettiker-Probe model but takes into account all relevant individual scattering mechanisms. It is orders of magnitude more efficient than a fully selfconsistent non-equilibrium Green's function calculation for realistic devices, yet accurately reproduces the results of the latter method as well as experimental data. This method is fairly easy to implement and opens the path towards realistic three-dimensional quantum transport calculations. In this work, we review the fundamentals of the non-equilibrium Green's function formalism for quantum transport calculations. Then, we introduce our novel MSB method after briefly reviewing the original Buettiker-Probe model. Finally, we compare the results of the MSB method to NEGF calculations as well as to experimental data. In particular, we calculate quantum transport properties of quantum cascade lasers in the terahertz (THz) and the mid-infrared (MIR) spectral domain. With a device optimization algorithm based upon the MSB method, we propose a novel THz quantum cascade laser design. It uses a two-well period with alternating barrier heights and complete carrier thermalization for the majority of the carriers within each period. We predict THz laser operation for temperatures up to 250 K implying a new temperature record.

A Methodology for Physical Interconnection Decisions of Next Generation Transport Networks

DEFF Research Database (Denmark)

Gutierrez Lopez, Jose Manuel; Riaz, M. Tahir; Madsen, Ole Brun

2011-01-01

of possibilities when designing the physical network interconnection. This paper develops and presents a methodology in order to deal with aspects related to the interconnection problem of optical transport networks. This methodology is presented as independent puzzle pieces, covering diverse topics going from......The physical interconnection for optical transport networks has critical relevance in the overall network performance and deployment costs. As telecommunication services and technologies evolve, the provisioning of higher capacity and reliability levels is becoming essential for the proper...... development of Next Generation Networks. Currently, there is a lack of specific procedures that describe the basic guidelines to design such networks better than "best possible performance for the lowest investment". Therefore, the research from different points of view will allow a broader space...

Validating analysis methodologies used in burnup credit criticality calculations

International Nuclear Information System (INIS)

Brady, M.C.; Napolitano, D.G.

1992-01-01

The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

Implementation and adaptation of a macro-scale methodology to calculate direct economic losses

Science.gov (United States)

Natho, Stephanie; Thieken, Annegret

2017-04-01

As one of the 195 member countries of the United Nations, Germany signed the Sendai Framework for Disaster Risk Reduction 2015-2030 (SFDRR). With this, though voluntary and non-binding, Germany agreed to report on achievements to reduce disaster impacts. Among other targets, the SFDRR aims at reducing direct economic losses in relation to the global gross domestic product by 2030 - but how to measure this without a standardized approach? The United Nations Office for Disaster Risk Reduction (UNISDR) has hence proposed a methodology to estimate direct economic losses per event and country on the basis of the number of damaged or destroyed items in different sectors. The method bases on experiences from developing countries. However, its applicability in industrial countries has not been investigated so far. Therefore, this study presents the first implementation of this approach in Germany to test its applicability for the costliest natural hazards and suggests adaptations. The approach proposed by UNISDR considers assets in the sectors agriculture, industry, commerce, housing, and infrastructure by considering roads, medical and educational facilities. The asset values are estimated on the basis of sector and event specific number of affected items, sector specific mean sizes per item, their standardized construction costs per square meter and a loss ratio of 25%. The methodology was tested for the three costliest natural hazard types in Germany, i.e. floods, storms and hail storms, considering 13 case studies on the federal or state scale between 1984 and 2016. Not any complete calculation of all sectors necessary to describe the total direct economic loss was possible due to incomplete documentation. Therefore, the method was tested sector-wise. Three new modules were developed to better adapt this methodology to German conditions covering private transport (cars), forestry and paved roads. Unpaved roads in contrast were integrated into the agricultural and

A meshless approach to radionuclide transport calculations

International Nuclear Information System (INIS)

Perko, J.; Sarler, B.

2005-01-01

Over the past thirty years numerical modelling has emerged as an interdisciplinary scientific discipline which has a significant impact in engineering and design. In the field of numerical modelling of transport phenomena in porous media, many commercial codes exist, based on different numerical methods. Some of them are widely used for performance assessment and safety analysis of radioactive waste repositories and groundwater modelling. Although they proved to be an accurate and reliable tool, they have certain limitations and drawbacks. Realistic problems often involve complex geometry which is difficult and time consuming to discretize. In recent years, meshless methods have attracted much attention due to their flexibility in solving engineering and scientific problems. In meshless methods the cumbersome polygonization of calculation domain is not necessary. By this the discretization time is reduced. In addition, the simulation is not as discretization density dependent as in traditional methods because of the lack of polygon interfaces. In this work fully meshless Diffuse Approximate Method (DAM) is used for calculation of radionuclide transport. Two cases are considered; First 1D comparison of 226 Ra transport and decay solved by the commercial Finite Volume Method (FVM) and Finite Element Method (FEM) based packages and DAM. This case shows the level of discretization density dependence. And second realistic 2D case of near-field modelling of radionuclide transport from the radioactive waste repository. Comparison is made again between FVM based code and DAM simulation for two radionuclides: Long-lived 14 C and short-lived 3 H. Comparisons indicate great capability of meshless methods to simulate complex transport problems and show that they should be seriously considered in future commercial simulation tools. (author)

Neutron transport calculations of some fast critical assemblies

Energy Technology Data Exchange (ETDEWEB)

Martinez-Val Penalosa, J A

1976-07-01

To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.

Neutron transport calculations of some fast critical assemblies

International Nuclear Information System (INIS)

Martinez-Val Penalosa, J. A.

1976-01-01

To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs

CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

Energy Technology Data Exchange (ETDEWEB)

Vinson, D.

2010-07-11

Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and

Resonance self-shielding methodology of new neutron transport code STREAM

International Nuclear Information System (INIS)

Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi

2015-01-01

This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)

Calculation of transportation energy for biomass collection

Energy Technology Data Exchange (ETDEWEB)

Kanai, G.; Takekura, K.; Kato, H.; Kobayashi, Y.; Yakushido, K. [National Agricultural Research Center, Tsukuba, Ibaraki (Japan)

2010-07-01

This paper reported on a study at a rice straw facility in Japan that produces bioethanol. Simulation modeling and calculations methods were used to examine the characteristics of field-to-facility transportation. Fuel consumption was found to be influenced by the conversion rate from straw to ethanol, the quantity of straw collected, and the ratio of the field area to that around the facility. Standard conditions were assumed based on reported data and actual observations for 15 ML/yr ethanol production, 0.3 kL output of ethanol from 1 t dry straw, 53.6 day/yr working days, 2.7 t truck load capacity, and 0.128 as the ratio of field to the area around the facility. According to calculations, a quantity of 50 kt dry straw requires 2.78 L of fuel to transport 1 t of dry straw, 109.5 trucks, and a 19.1 km collection area radius. The fuel consumption for transportation was found to be proportional to the quantity of straw to the 0.5 power, but inversely proportional to the ratio of field to the 0.5 power. The rate of increase in the number of trucks needed to collect straw increases with the decrease in the ratio of the field to area surface around the facility.

Whole core transport calculation for the VHTR hexagonal core

International Nuclear Information System (INIS)

Cho, J. Y.; Kim, K. S.; Lee, C. C.; Joo, H. G.

2007-01-01

Recently, the DeCART code which performs the whole core calculation by coupling the radial MOC transport kernel with the axial nodal kernel has equipped a kernel to deal with the hexagonal geometry and applied to the VHTR hexagonal core to examine the accuracy and the computational efficiency of the implemented kernel. The implementation includes a modular ray tracing module based on the hexagonal assembly and a multi-group CMFD module to perform an efficient transport calculation. The requirements for the modular ray are: (1) the assembly based path linking and (2) the complete reflection capabilities. The first requirement is met by adjusting the azimuthal angle and the ray spacing for the modular ray to construct a core ray by the path linking. The second requirement is met by expanding the constructed azimuthal angle in the range of [0,30 degree] to the remained range to reflect completely at the core boundaries. The considered reflecting surface angles for the complete reflection are 30n's (n=1,2,1,12). The CMFD module performs the equivalent diffusion calculation to the radial MOC transport calculation based on the homogenized structure units. The structure units include the hexagonal pin cells and gap cells appearing at the assembly boundary. Therefore, the CMFD module is programmed to deal with the unstructured cells such as the gap cells. The CMFD equation consists of the two parts of (1) the conventional FDM and (2) the current corrective parts. Since the second part of the CMFD equation guarantees the reproducibility of the radial MOC transport solutions for the cell averaged reaction rate and the net current at the cell surfaces, how to build the first part of the CMFD equation is not important. Therefore, the first part of the CMFD equation is roughly built by using the normal distance from the gravity center to the surface. The VHTR core uses helium as a coolant which is realized as a void hole in a neutronics calculation. This void hole which

Optimal calculational schemes for solving multigroup photon transport problem

International Nuclear Information System (INIS)

Dubinin, A.A.; Kurachenko, Yu.A.

1987-01-01

A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

Methodology for coupling computational fluid dynamics and integral transport neutronics

International Nuclear Information System (INIS)

Thomas, J. W.; Zhong, Z.; Sofu, T.; Downar, T. J.

2004-01-01

The CFD code STAR-CD was coupled to the integral transport code DeCART in order to provide high-fidelity, full physics reactor simulations. An interface program was developed to perform the tasks of mapping the STAR-CD mesh to the DeCART mesh, managing all communication between STAR-CD and DeCART, and monitoring the convergence of the coupled calculations. The interface software was validated by comparing coupled calculation results with those obtained using an independently developed interface program. An investigation into the convergence characteristics of coupled calculations was performed using several test models on a multiprocessor LINUX cluster. The results indicate that the optimal convergence of the coupled field calculation depends on several factors, to include the tolerance of the STAR-CD solution and the number of DeCART transport sweeps performed before exchanging data between codes. Results for a 3D, multi-assembly PWR problem on 12 PEs of the LINUX cluster indicate the best performance is achieved when the STAR-CD tolerance and number of DeCART transport sweeps are chosen such that the two fields converge at approximately the same rate. (authors)

Parallel processing of neutron transport in fuel assembly calculation

International Nuclear Information System (INIS)

Song, Jae Seung

1992-02-01

Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

Microwave emulations and tight-binding calculations of transport in polyacetylene

Energy Technology Data Exchange (ETDEWEB)

Stegmann, Thomas, E-mail: stegmann@icf.unam.mx [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Franco-Villafañe, John A., E-mail: jofravil@fis.unam.mx [Instituto de Física, Benemérita Universidad Autónoma de Puebla, Apartado Postal J-48, 72570 Puebla (Mexico); Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Ortiz, Yenni P. [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Kuhl, Ulrich [Université de Nice – Sophia Antipolis, Laboratoire de la Physique de la Matière Condensée, CNRS, Parc Valrose, 06108 Nice (France); Mortessagne, Fabrice, E-mail: fabrice.mortessagne@unice.fr [Université de Nice – Sophia Antipolis, Laboratoire de la Physique de la Matière Condensée, CNRS, Parc Valrose, 06108 Nice (France); Seligman, Thomas H. [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Centro Internacional de Ciencias, 62210 Cuernavaca (Mexico)

2017-01-05

A novel approach to investigate the electron transport of cis- and trans-polyacetylene chains in the single-electron approximation is presented by using microwave emulation measurements and tight-binding calculations. In the emulation we take into account the different electronic couplings due to the double bonds leading to coupled dimer chains. The relative coupling constants are adjusted by DFT calculations. For sufficiently long chains a transport band gap is observed if the double bonds are present, whereas for identical couplings no band gap opens. The band gap can be observed also in relatively short chains, if additional edge atoms are absent, which cause strong resonance peaks within the band gap. The experimental results are in agreement with our tight-binding calculations using the nonequilibrium Green's function method. The tight-binding calculations show that it is crucial to include third nearest neighbor couplings to obtain the gap in the cis-polyacetylene. - Highlights: • Electronic transport in individual polyacetylene chains is studied. • Microwave emulation experiments and tight-binding calculations agree well. • In long chains a band-gap opens due the dimerization of the chain. • In short chains edge atoms cause strong resonance peaks in the center of the band-gap.

Microwave emulations and tight-binding calculations of transport in polyacetylene

International Nuclear Information System (INIS)

Stegmann, Thomas; Franco-Villafañe, John A.; Ortiz, Yenni P.; Kuhl, Ulrich; Mortessagne, Fabrice; Seligman, Thomas H.

2017-01-01

A novel approach to investigate the electron transport of cis- and trans-polyacetylene chains in the single-electron approximation is presented by using microwave emulation measurements and tight-binding calculations. In the emulation we take into account the different electronic couplings due to the double bonds leading to coupled dimer chains. The relative coupling constants are adjusted by DFT calculations. For sufficiently long chains a transport band gap is observed if the double bonds are present, whereas for identical couplings no band gap opens. The band gap can be observed also in relatively short chains, if additional edge atoms are absent, which cause strong resonance peaks within the band gap. The experimental results are in agreement with our tight-binding calculations using the nonequilibrium Green's function method. The tight-binding calculations show that it is crucial to include third nearest neighbor couplings to obtain the gap in the cis-polyacetylene. - Highlights: • Electronic transport in individual polyacetylene chains is studied. • Microwave emulation experiments and tight-binding calculations agree well. • In long chains a band-gap opens due the dimerization of the chain. • In short chains edge atoms cause strong resonance peaks in the center of the band-gap.

Life Cycle Cost Calculation at the Transport Company in the Supply of Production of Wooden Houses – Case Study

Directory of Open Access Journals (Sweden)

Potkány Marek

2017-01-01

Full Text Available A correct information manager's decision-maker database is a very important element that substantially affects its success. This article presents the potential of using the methodology of life cycle cost calculation in the conditions of a transport company that focuses on the logistic supply of wood-housing producers. The problem is presented through a case study and addresses the decision-making aspect of the decision about acquisition of the transport vehicle. This decision uses time value indicators, inflation rates, average rate of profitability of industry and life cycle costs. Due to the short life cycle of the analyzed period, it was not necessary to consider the ergonomic requirements resulting from the trend of anthropometric dimensions growth.

Load Balancing of Parallel Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

Procassini, R J; O'Brien, M J; Taylor, J M

2005-01-01

The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since he particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

Discussion of electron cross sections for transport calculations

International Nuclear Information System (INIS)

Berger, M.J.

1983-01-01

This paper deals with selected aspects of the cross sections needed as input for transport calculations and for the modeling of radiation effects in biological materials. Attention is centered mainly on the cross sections for inelastic interactions between electrons and water molecules and the use of these cross sections for the calculation of energy degradation spectra and of ionization and excitation yields. 40 references, 3 figures, 1 table

A calculational methodology for comparing the accident, occupational, and waste-disposal hazards of fusion reactor designs

International Nuclear Information System (INIS)

Fetter, S.

1985-01-01

A methodology has been developed for calculating indices of three classes of radiological hazards: reactor accidents, occupational exposures, and waste-disposal hazards. Radionuclide inventories, biological hazard potentials (BHP), and various dose-related indices are calculated. In the case of reactor accidents, the critical, 50-year and chronic dose are computed, as well as the number of early deaths and illnesses and late cancer fatalities. For occupational exposure, the contact dose rate is calculated for several times after reactor shutdown. In the case of waste-disposal hazards, the intruder dose and the intruder hazard potential (IHP) are calculated. Sample calculations for the MARS reactor design show the usefulness of the methodology in exploring design improvements

Calculation and evaluation methodology of the flawed pipe and the compute program development

International Nuclear Information System (INIS)

Liu Chang; Qian Hao; Yao Weida; Liang Xingyun

2013-01-01

Background: The crack will grow gradually under alternating load for a pressurized pipe, whereas the load is less than the fatigue strength limit. Purpose: Both calculation and evaluation methodology for a flawed pipe that have been detected during in-service inspection is elaborated here base on the Elastic Plastic Fracture Mechanics (EPFM) criteria. Methods: In the compute, the depth and length interaction of a flaw has been considered and a compute program is developed per Visual C++. Results: The fluctuating load of the Reactor Coolant System transients, the initial flaw shape, the initial flaw orientation are all accounted here. Conclusions: The calculation and evaluation methodology here is an important basis for continue working or not. (authors)

Application of the adjoint function methodology for neutron fluence determination

International Nuclear Information System (INIS)

Haghighat, A.; Nanayakkara, B.; Livingston, J.; Mahgerefteh, M.; Luoma, J.

1991-01-01

In previous studies, the neutron fluence at a reactor pressure vessel has been estimated based on consolidation of transport theory calculations and experimental data obtained from in-vessel capsules and/or cavity dosimeters. Normally, a forward neutron transport calculation is performed for each fuel cycle and the neutron fluxes are integrated over the reactor operating time to estimate the neutron fluence. Such calculations are performed for a geometrical model which is composed of one-eighth (0 to 45 deg) of the reactor core and its surroundings; i.e., core barrel, thermal shield, downcomer, reactor vessel, cavity region, concrete wall, and instrumentation well. Because the model is large, transport theory calculations generally require a significant amount of computer memory and time; hence, more efficient methodologies such as the adjoint transport approach have been proposed. These studies, however, do not address the necessary sensitivity studies needed for adjoint function calculations. The adjoint methodology has been employed to estimate the activity of a cavity dosimeter and that of an in-vessel capsule. A sensitivity study has been performed on the mesh distribution used in and around the cavity dosimeter and the in-vessel capsule. Further, since a major portion of the detector response is due to the neutrons originated in the peripheral fuel assemblies, a study on the use of a smaller calculational model has been performed

Uncertainty analysis of neutron transport calculation

International Nuclear Information System (INIS)

Oka, Y.; Furuta, K.; Kondo, S.

1987-01-01

A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

LTRACK: Beam-transport calculation including wakefield effects

International Nuclear Information System (INIS)

Chan, K.C.D.; Cooper, R.K.

1988-01-01

LTRACK is a first-order beam-transport code that includes wakefield effects up to quadrupole modes. This paper will introduce the readers to this computer code by describing the history, the method of calculations, and a brief summary of the input/output information. Future plans for the code will also be described

Development of a calculation methodology for potential flow over irregular topographies

International Nuclear Information System (INIS)

Del Carmen, Alejandra F.; Ferreri, Juan C.; Boutet, Luis I.

2003-01-01

Full text: Computer codes for the calculation of potential flow fields over surfaces with irregular topographies have been developed. The flows past multiple simple obstacles and past the neighboring region of the Embalse Nuclear Power Station have been considered. The codes developed allow the calculation of velocities quite near the surface. It, in turn, imposed developing high accuracy techniques. The Boundary Element Method, using a linear approximation on triangular plane elements and an analytical integration methodology has been applied. A particular and quite efficient technique for the calculation of the solid angle at each node vertex was also considered. The results so obtained will be applied to predict the dispersion of passive pollutants coming from discontinuous emissions. (authors)

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

International Nuclear Information System (INIS)

White, Morgan C.

2000-01-01

The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

Energy Technology Data Exchange (ETDEWEB)

White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

2000-07-01

The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

Cost-effectiveness of greenhouse gas mitigation in transport: A review of methodological approaches and their impact

International Nuclear Information System (INIS)

Kok, Robert; Annema, Jan Anne; Wee, Bert van

2011-01-01

A review is given of methodological practices for ex ante cost-effectiveness analysis (CEA) of transport greenhouse gas (GHG) mitigation measures, e.g. fuel economy and CO 2 standards for road vehicles in the US and EU. Besides the fundamental differences between different types of policies and abatement options which inherently result in different CEA outcomes, differences in methodological choices and assumptions are another important source of variation in CEA outcomes. Fourteen methodological issues clustered into six groups are identified on which thirty-three selected studies are systematically reviewed. The potential variation between lower and upper cost-effectiveness estimates for GHG mitigation measures in transport, resulting from different methodological choices and assumptions, lies in the order of $400 per tonne CO 2 -eq. The practise of using CEA for policy-making could improve considerably by clearly indicating the specific purpose of the CEA and its strengths and limitations for policy decisions. Another improvement is related to the dominant approach in transport GHG mitigation studies: the bottom-up financial technical approach which assesses isolated effects, implying considerable limitations for policy-making. A shift to welfare-economic approaches using a hybrid model has the potential to establish an improved assessment of transport GHG mitigation measures based on realistic market responses and behavioural change. - Highlights: ► We identify fourteen important methodological issues clustered into six groups. ► We systematically review thirty-three selected transport GHG mitigation studies. ► Methodological choices can lead to a difference by up to $400 per tonne CO 2 -eq. ► The dominant bottom-up approach has limitations for policy-making. ► Welfare-economic approaches could improve cost-effectiveness analysis.

Impact limiters for radioactive materials transport packagings: a methodology for assessment

International Nuclear Information System (INIS)

Mourao, Rogerio Pimenta

2002-01-01

This work aims at establishing a methodology for design assessment of a cellular material-filled impact limiter to be used as part of a radioactive material transport packaging. This methodology comprises the selection of the cellular material, its structural characterization by mechanical tests, the development of a case study in the nuclear field, preliminary determination of the best cellular material density for the case study, performance of the case and its numerical simulation using the finite element method. Among the several materials used as shock absorbers in packagings, the polyurethane foam was chosen, particularly the foam obtained from the castor oil plant (Ricinus communis), a non-polluting and renewable source. The case study carried out was the 9 m drop test of a package prototype containing radioactive wastes incorporated in a cement matrix, considered one of the most severe tests prescribed by the Brazilian and international transport standards. Prototypes with foam density pre-determined as ideal as well as prototypes using lighter and heavier foams were tested for comparison. The results obtained validate the methodology in that expectations regarding the ideal foam density were confirmed by the drop tests and the numerical simulation. (author)

CALCULATING BEDLOAD TRANSPORT IN RIVERS: CONCEPTS, CALCULUS ROUTINES AND APPLICATION

Directory of Open Access Journals (Sweden)

Hudson de Azevedo Macedo

2017-10-01

Full Text Available Rivers are immensely important to human activities such as water supply, navigation, energy generation, and agriculture. They are also an important morphodynamic agent of erosion, transport and deposition. Their capacity to transport sediment depends on their hydraulic characteristics and can be predicted by mathematical models. Several mathematical models can be used to compute bed-load transport. Each one is appropriately better for certain conditions. In this paper, we present an application built in Microsoft Excel to compute the bed-load transport in rivers based on the Van Rijn mathematical model. The Van Rijn model is appropriate for rivers transporting sandy sediments in conditions of subcritical flow. Hydraulic parameters such as channel slope, stream power, and Reynolds and Froude numbers can be calculated using the application proposed here. The application was tested in the Paraná River and results from the calculations are consistent with data obtained from fieldwork surveys. The error of the application was only 20%, which shows good agreement of the model with survey values.

Parameters calculation of shielding experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-02-01

The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

Calculations of the transport properties within the PAW formalism

Energy Technology Data Exchange (ETDEWEB)

Mazevet, S.; Torrent, M.; Recoules, V.; Jollet, F. [CEA Bruyeres-le-Chatel, DIF, 91 (France)

2010-07-01

We implemented the calculation of the transport properties within the PAW formalism in the ABINIT code. This feature allows the calculation of the electrical and optical properties, including the XANES spectrum, as well as the electronic contribution to the thermal conductivity. We present here the details of the implementation and results obtained for warm dense aluminum plasma. (authors)

Synergism of the method of characteristics and CAD technology for neutron transport calculation

International Nuclear Information System (INIS)

Chen, Z.; Wang, D.; He, T.; Wang, G.; Zheng, H.

2013-01-01

The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)

Analysis of error in Monte Carlo transport calculations

International Nuclear Information System (INIS)

Booth, T.E.

1979-01-01

The Monte Carlo method for neutron transport calculations suffers, in part, because of the inherent statistical errors associated with the method. Without an estimate of these errors in advance of the calculation, it is difficult to decide what estimator and biasing scheme to use. Recently, integral equations have been derived that, when solved, predicted errors in Monte Carlo calculations in nonmultiplying media. The present work allows error prediction in nonanalog Monte Carlo calculations of multiplying systems, even when supercritical. Nonanalog techniques such as biased kernels, particle splitting, and Russian Roulette are incorporated. Equations derived here allow prediction of how much a specific variance reduction technique reduces the number of histories required, to be weighed against the change in time required for calculation of each history. 1 figure, 1 table

Contribution to gamma ray transport calculation in heterogeneous media

International Nuclear Information System (INIS)

Bourdet, L.

1985-04-01

This thesis presents the development of gamma transport calculation codes in three dimension heterogeneous geometries. These codes allow us to define the protection against gamma-rays or verify their efficiency. The laws that govern the interactions of gamma-rays with matters are briefly revised. A library with the all necessary constants for these codes is created. TRIPOLI-2, a code that treats in exact way the neutron transport in matters using Monte-Carlo method, has been adapted to deal with the transport of gamma-rays in matters as well. TRINISHI, a code which considers only one collision, has been realized to treat heterogeneous geometries containing voids. Elaborating a formula that calculates the albedo for gamma-ray reflection (the code ALBANE) allows us to solve the problem of gamma-ray reflection on plane surfaces. NARCISSE-2 deals with gamma-rays that suffer only one reflection on the inner walls of any closed volume (rooms, halls...) [fr

HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

International Nuclear Information System (INIS)

Honeck, H.C.

1984-01-01

1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

Transport calculation of medium-energy protons and neutrons by Monte Carlo method

International Nuclear Information System (INIS)

Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.

1978-09-01

A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)

Lagrangian Transport Calculations Using UARS Data. Part 2; Ozone

Science.gov (United States)

Manney, Gloria L.; Zurek, R. W.; Froidevaux, L.; Waters, J. W.; ONeill, A.; Swinbank, R.

1995-01-01

Trajectory calculations are used to examine ozone transport in the polar winter stratosphere during periods of the Upper Atmosphere Research Satellite (UARS) observations. The value of these calculations for determining mass transport was demonstrated previously using UARS observations of long-lived tracers, In the middle stratosphere, the overall ozone behavior observed by the Microwave Limb Sounder in the polar vortex is reproduced by this purely dynamical model. Calculations show the evolution of ozone in the lower stratosphere during early winter to be dominated by dynamics in December 1992 in the Arctic. Calculations for June 1992 in the Antarctic show evidence of chemical ozone destruction and indicate that approx. 50% of the chemical destruction may be masked by dynamical effects, mainly diabatic descent, which bring higher ozone into the lower-stratospheric vortex. Estimating differences between calculated and observed fields suggests that dynamical changes masked approx. 20% - 35% of chemical ozone loss during late February and early March 1993 in the Arctic. In the Antarctic late winter, in late August and early September 1992, below approx. 520 K, the evolution of vortex-averaged ozone is entirely dominated by chemical effects; above this level, however, chemical ozone depletion can be partially or completely masked by dynamical effects. Our calculations for 1992 showed that chemical loss was nearly completely compensated by increases due to diabatic descent at 655 K.

Modeling Dynamic Objects in Monte Carlo Particle Transport Calculations

International Nuclear Information System (INIS)

Yegin, G.

2008-01-01

In this study, the Multi-Geometry geometry modeling technique was improved in order to handle moving objects in a Monte Carlo particle transport calculation. In the Multi-Geometry technique, the geometry is a superposition of objects not surfaces. By using this feature, we developed a new algorithm which allows a user to make enable or disable geometry elements during particle transport. A disabled object can be ignored at a certain stage of a calculation and switching among identical copies of the same object located adjacent poins during a particle simulation corresponds to the movement of that object in space. We called this powerfull feature as Dynamic Multi-Geometry technique (DMG) which is used for the first time in Brachy Dose Monte Carlo code to simulate HDR brachytherapy treatment systems. Our results showed that having disabled objects in a geometry does not effect calculated dose values. This technique is also suitable to be used in other areas such as IMRT treatment planning systems

Photonuclear Physics in Radiation Transport - II: Implementation

International Nuclear Information System (INIS)

White, M.C.; Little, R.C.; Chadwick, M.B.; Young, P.G.; MacFarlane, R.E.

2003-01-01

This is the second of two companion papers. The first paper describes model calculations and nuclear data evaluations of photonuclear reactions on isotopes of C, O, Al, Si, Ca, Fe, Cu, Ta, W, and Pb for incident photon energies up to 150 MeV. This paper describes the steps taken to process these files into transport libraries and to update the Monte Carlo N-Particle (MCNP) and MCNPX radiation transport codes to use tabular photonuclear reaction data. The evaluated photonuclear data files are created in the standard evaluated nuclear data file (ENDF) format. These files must be processed by the NJOY data processing system into A Compact ENDF (ACE) files suitable for radiation transport calculations. MCNP and MCNPX have been modified to use these new data in a self-consistent and fully integrated manner. Verification problems were used at each step along the path to check the integrity of the methodology. The resulting methodology and tools provide a comprehensive system for using photonuclear data in radiation transport calculations. Also described are initial validation simulations used to benchmark several of the photonuclear transport tables

Transportation channels calculation method in MATLAB

International Nuclear Information System (INIS)

Averyanov, G.P.; Budkin, V.A.; Dmitrieva, V.V.; Osadchuk, I.O.; Bashmakov, Yu.A.

2014-01-01

Output devices and charged particles transport channels are necessary components of any modern particle accelerator. They differ both in sizes and in terms of focusing elements depending on particle accelerator type and its destination. A package of transport line designing codes for magnet optical channels in MATLAB environment is presented in this report. Charged particles dynamics in a focusing channel can be studied easily by means of the matrix technique. MATLAB usage is convenient because its information objects are matrixes. MATLAB allows the use the modular principle to build the software package. Program blocks are small in size and easy to use. They can be executed separately or commonly. A set of codes has a user-friendly interface. Transport channel construction consists of focusing lenses (doublets and triplets). The main of the magneto-optical channel parameters are total length and lens position and parameters of the output beam in the phase space (channel acceptance, beam emittance - beam transverse dimensions, particles divergence and image stigmaticity). Choice of the channel operation parameters is based on the conditions for satisfying mutually competing demands. And therefore the channel parameters calculation is carried out by using the search engine optimization techniques.

Calculations on safe storage and transportation of radioactive materials

Energy Technology Data Exchange (ETDEWEB)

Hathout, A M; El-Messiry, A M; Amin, E [National Center for Nuclear Safety and Radiation Control and AEA, Cairo (Egypt)

1997-12-31

In this work the safe storage and transportation of fresh fuel as a radioactive material studied. Egypt planned ET RR 2 reactor which is of relatively high power and would require adequate handling and transportation. Therefore, the present work is initiated to develop a procedure for safe handling and transportation of radioactive materials. The possibility of reducing the magnitude of radiation transmitted on the exterior of the packages is investigated. Neutron absorbers are used to decrease the neutron flux. Criticality calculations are carried out to ensure the achievement of subcriticality so that the inherent safety can be verified. The discrete ordinate transport code ANISN was used. The results show good agreement with other techniques. 2 figs., 2 tabs.

Accounting for chemical kinetics in field scale transport calculations

International Nuclear Information System (INIS)

Bryan, N.D.

2005-01-01

The modelling of column experiments has shown that the humic acid mediated transport of metal ions is dominated by the non-exchangeable fraction. Metal ions enter this fraction via the exchangeable fraction, and may transfer back again. However, in both directions these chemical reactions are slow. Whether or not a kinetic description of these processes is required during transport calculations, or an assumption of local equilibrium will suffice, will depend upon the ratio of the reaction half-time to the residence time of species within the groundwater column. If the flow rate is sufficiently slow or the reaction sufficiently fast then the assumption of local equilibrium is acceptable. Alternatively, if the reaction is sufficiently slow (or the flow rate fast), then the reaction may be 'decoupled', i.e. removed from the calculation. These distinctions are important, because calculations involving chemical kinetics are computationally very expensive, and should be avoided wherever possible. In addition, column experiments have shown that the sorption of humic substances and metal-humate complexes may be significant, and that these reactions may also be slow. In this work, a set of rules is presented that dictate when the local equilibrium and decoupled assumptions may be used. In addition, it is shown that in all cases to a first approximation, the behaviour of a kinetically controlled species, and in particular its final distribution against distance at the end of a calculation, depends only upon the ratio of the reaction first order rate to the residence time, and hence, even in the region where the simplifications may not be used, the behaviour is predictable. In this way, it is possible to obtain an estimate of the migration of these species, without the need for a complex transport calculation. (orig.)

Methodology of personnel exposure calculation and optimisation within the decommissioning planning code OMEGA

International Nuclear Information System (INIS)

Vasko, Marek; Daniska, Vladimir; Rehak, Ivan; Necas, Vladimir

2011-01-01

Calculation of personnel exposure is a one of the main parameters being evaluated within the pre-decommissioning plans together with other decommissioning drivers such as costs, manpower, amounts of RAW and conventional waste and amount of discharged gaseous and liquid effluents. Alongside with manpower, the exposure is an indicator of the decommissioning process for need of staff, and quantifies impact of decommissioning on personnel from the radio hygienic point of view. At the same time it indicates suitability of individual work procedures use for decommissioning activities. For this reason it is important to estimate as precise as possible demands on personnel exposure even during preparatory decommissioning phase to quantify impact of decommissioning on personnel and eventually optimize the decommissioning process, if needed. The most appropriate way of staff exposure estimation during decommissioning preparatory phases is its calculation based on radiological and physical characteristics of equipment to be decommissioned and also quantitative and qualitative characterisation of typical decommissioning activities. On one hand, the methodology of exposure calculation should allow as much as possible realistic description and algorithmisation of exposure ways during decommissioning activities. On the other hand the calculation have to be systematic, well-arranged and clearly definable by appropriate mathematic relations. Calculation can be made by various approaches using more or less sophisticated software solutions from classic MS Excel sheets up to the complex calculation codes. In this paper, a methodology used for personnel exposure calculation and optimization implemented within the complex computer code OMEGA developed at DECOM, a.s. is described. (author)

Uniform Gauss-Weight Quadratures for Discrete Ordinate Transport Calculations

International Nuclear Information System (INIS)

Carew, John F.; Hu, Kai; Zamonsky, Gabriel

2000-01-01

Recently, a uniform equal-weight quadrature set, UE n , and a uniform Gauss-weight quadrature set, UG n , have been derived. These quadratures have the advantage over the standard level-symmetric LQ n quadrature sets in that the weights are positive for all orders,and the transport solution may be systematically converged by increasing the order of the quadrature set. As the order of the quadrature is increased,the points approach a uniform continuous distribution on the unit sphere,and the quadrature is invariant with respect to spatial rotations. The numerical integrals converge for continuous functions as the order of the quadrature is increased.The numerical characteristics of the UE n quadrature set have been investigated previously. In this paper, numerical calculations are performed to evaluate the application of the UG n quadrature set in typical transport analyses. A series of DORT transport calculations of the >1-MeV neutron flux have been performed for a set of pressure-vessel fluence benchmark problems. These calculations employed the UG n (n = 8, 12, 16, 24, and 32) quadratures and indicate that the UG n solutions have converged to within ∼0.25%. The converged UG n solutions are found to be comparable to the UE n results and are more accurate than the level-symmetric S 16 predictions

Evaluation of dose equivalent rate distribution in JCO critical accident by radiation transport calculation

CERN Document Server

Sakamoto, Y

2002-01-01

In the prevention of nuclear disaster, there needs the information on the dose equivalent rate distribution inside and outside the site, and energy spectra. The three dimensional radiation transport calculation code is a useful tool for the site specific detailed analysis with the consideration of facility structures. It is important in the prediction of individual doses in the future countermeasure that the reliability of the evaluation methods of dose equivalent rate distribution and energy spectra by using of Monte Carlo radiation transport calculation code, and the factors which influence the dose equivalent rate distribution outside the site are confirmed. The reliability of radiation transport calculation code and the influence factors of dose equivalent rate distribution were examined through the analyses of critical accident at JCO's uranium processing plant occurred on September 30, 1999. The radiation transport calculations including the burn-up calculations were done by using of the structural info...

Dynamic Load Balancing of Parallel Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

O'Brien, M; Taylor, J; Procassini, R

2004-01-01

The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since the particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

A retrospective and prospective survey of three-dimensional transport calculations

International Nuclear Information System (INIS)

Nakahara, Yasuaki

1985-01-01

A retrospective survey is made on the three-dimensional radiation transport calculations. Introduction is given to computer codes based on the distinctive numerical methods such as the Monte Carlo, Direct Integration, Ssub(n) and Finite Element Methods to solve the three-dimensional transport equations. Prospective discussions are made on pros and cons of these methods. (author)

Methodology for a thermal analysis of a proposed SFR transport cask with the thermal code SYRTHES

International Nuclear Information System (INIS)

Peniguel, C.; Rupp, I.; Schneider, J. P.

2010-01-01

Fast reactors with liquid metal coolant have received a renewed interest owing to the need of a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of minor actinides transmutation advantages and drawbacks in Generation IV fast spectrum reactors system is requested for 2012. The CEA is in charge of studying the global problem, but on some aspects, EDF is interested to do its own exploratory studies. Among other points, transport is seen as important for the nuclear industry, to link points of production and treatment. Nuclear fuel is generally transported in thick walled rail or truck casks. These packages are designed to provide confinement, shielding and criticality protection during normal and severe transport conditions. Heat generated within the fuel (and a contribution of solar heating) makes the package becoming quite hot, but one must demonstrate that the cladding temperature does not exceed a long term temperature limit during normal transport. This paper presents a thermal study done on a package in which 9 SFR assemblies are included. Each of them is of hexagonal shape and contains 271 fuel pins. The approach followed for these calculations is to rely on an explicit representation of all pins. For these calculations a 2D analysis is performed thanks to the thermal code SYRTHES. Conduction is solved thanks to a finite element method, while thermal radiation is handled through a radiosity approach. The main aim of this paper is to present a possible numerical methodology to handle the thermal problem. (authors)

Spectral nodal methodology for multigroup slab-geometry discrete ordinates neutron transport problems with linearly anisotropic scattering

Energy Technology Data Exchange (ETDEWEB)

Oliva, Amaury M.; Filho, Hermes A.; Silva, Davi M.; Garcia, Carlos R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: davijmsilva@yahoo.com.br, E-mail: cgh@instec.cu [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional; Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

2017-07-01

In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that will generate numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as an initial study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S{sub N}) formulation, in one-dimensional slab geometry, multigroup approximation, with linearly anisotropic scattering, considering homogeneous and heterogeneous domains with fixed source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional ne- mesh method Diamond Difference (DD) and the coarse-mesh method spectral Green's function (SGF) to illustrate the method's accuracy and stability. The solution algorithms problems are implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)

Methods for U.S. shielding calculations: applications to FFTF and CRBR designs

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Mynatt, F.R.; Disney, R.K.

1978-01-01

The primary components of the U.S. reactor shielding methodology consist of: (1) computer code systems based on discrete ordinates or Monte Carlo radiation transport calculational methods; (2) a data base of neutron and gamma-ray interaction and gamma-ray-production cross sections used as input in the codes; (3) a capability for processing the cross sections into multigroup or point energy formats as required by the codes; (4) large-scale integral shielding experiments designed to test cross-section data or techniques utilized in the calculations; and (5) a ''sensitivity'' analysis capability that can identify the most important interactions in a transport calculation and assign uncertainties to the calculated result that are based on uncertainties in all of the input data. The required accuracy for the methodology is to within 5 to 10% for responses at locations near the core to within a factor of 2 for responses at distant locations. Under these criteria, the methodology has proved to be adequate for in-vessel LMFBR calculations of neutron transport through deep sodium and thick iron and stainless steel shields, of neutron streaming through lower axial coolant channels and primary pipe chaseways, and of the effects of fuel stored within the reactor vessel. For ex-vessel LMFBR problems, the methodology requires considerable improvement, the areas of concern including neutron streaming through heating and ventilation ducts, through the cavity surrounding the reactor vessel, and through gaps around rotating plugs in the reactor heat, as well as gamma-ray streaming through plant shield penetrations

Methodological advances in unit cost calculation of psychiatric residential care in Spain.

Science.gov (United States)

Moreno, Karen; Sanchez, Eduardo; Salvador-Carulla, Luis

2008-06-01

The care of the severe mentally ill who need intensive support for their daily living (dependent persons), accounts for an increasingly large proportion of public expenditure in many European countries. The main aim of this study was the design and implementation of solid methodology to calculate unit costs of different types of care. To date, methodologies used in Spain have produced inaccurate figures, suggesting few variations in patient consumption of the same service. An adaptation of the Activity-Based-Costing methodology was applied in Navarre, a region in the North of Spain, as a pilot project for the public mental health services. A unit cost per care process was obtained for all levels of care considered in each service during 2005. The European Service Mapping Schedule (ESMS) codes were used to classify the services for later comparisons. Finally, in order to avoid problems of asymmetric cost distribution, a simple Bayesian model was used. As an illustration, we report the results obtained for long-term residential care and note that there are important variations between unit costs when considering different levels of care. Considering three levels of care (Level 1-low, Level 2-medium and Level 3-intensive), the cost per bed in Level 3 was 10% higher than that of Level 2. The results obtained using the cost methodology described provide more useful information than those using conventional methods, although its implementation requires much time to compile the necessary information during the initial stages and the collaboration of staff and managers working in the services. However, in some services, if no important variations exist in patient care, another method would be advisable, although our system provides very useful information about patterns of care from a clinical point of view. Detailed work is required at the beginning of the implementation in order to avoid the calculation of distorted figures and to improve the levels of decision making

Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

International Nuclear Information System (INIS)

Lefvert, T.

1975-11-01

Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

New methodologies for calculation of flight parameters on reduced scale wings models in wind tunnel =

Science.gov (United States)

Ben Mosbah, Abdallah

In order to improve the qualities of wind tunnel tests, and the tools used to perform aerodynamic tests on aircraft wings in the wind tunnel, new methodologies were developed and tested on rigid and flexible wings models. A flexible wing concept is consists in replacing a portion (lower and/or upper) of the skin with another flexible portion whose shape can be changed using an actuation system installed inside of the wing. The main purpose of this concept is to improve the aerodynamic performance of the aircraft, and especially to reduce the fuel consumption of the airplane. Numerical and experimental analyses were conducted to develop and test the methodologies proposed in this thesis. To control the flow inside the test sections of the Price-Paidoussis wind tunnel of LARCASE, numerical and experimental analyses were performed. Computational fluid dynamics calculations have been made in order to obtain a database used to develop a new hybrid methodology for wind tunnel calibration. This approach allows controlling the flow in the test section of the Price-Paidoussis wind tunnel. For the fast determination of aerodynamic parameters, new hybrid methodologies were proposed. These methodologies were used to control flight parameters by the calculation of the drag, lift and pitching moment coefficients and by the calculation of the pressure distribution around an airfoil. These aerodynamic coefficients were calculated from the known airflow conditions such as angles of attack, the mach and the Reynolds numbers. In order to modify the shape of the wing skin, electric actuators were installed inside the wing to get the desired shape. These deformations provide optimal profiles according to different flight conditions in order to reduce the fuel consumption. A controller based on neural networks was implemented to obtain desired displacement actuators. A metaheuristic algorithm was used in hybridization with neural networks, and support vector machine approaches and their

An integrated methodology for characterizing flow and transport processes in fractured rock

International Nuclear Information System (INIS)

Wu, Yu-Shu

2007-01-01

To investigate the coupled processes involved in fluid and heat flow and chemical transport in the highly heterogeneous, unsaturated-zone (UZ) fractured rock of Yucca Mountain, we present an integrated modeling methodology. This approach integrates a wide variety of moisture, pneumatic, thermal, and geochemical isotopic field data into a comprehensive three-dimensional numerical model for modeling analyses. The results of field applications of the methodology show that moisture data, such as water potential and liquid saturation, are not sufficient to determine in situ percolation flux, whereas temperature and geochemical isotopic data provide better constraints to net infiltration rates and flow patterns. In addition, pneumatic data are found to be extremely valuable in estimating large-scale fracture permeability. The integration of hydrologic, pneumatic, temperature, and geochemical data into modeling analyses is thereby demonstrated to provide a practical modeling approach for characterizing flow and transport processes in complex fractured formations

Methodology of calculation in one-dimensional kinetic

International Nuclear Information System (INIS)

Paixao, S.B.; Marzo, M.A.S.; Alvim, A.C.M.

1986-01-01

This paper resulted from a study of the WIGLE's program calculation method ]1], which is RESTRICTED to USA users. In view of this fact, a successful attempt was made to fully understand and reproduce the WIGLE methodology, thus providing support for national development on the subject. After finishing the theoretical study, CITER-1D, a program for search of control rod position in PWR slabs under steady-state conditions was written and is supposed to correctly reproduce WIGL3 ]4] version behavior. Program restriction to steady-state conditions was due to scarcity of examples, thought to be intentional, as well as to time limitations for conclusion of a M.Sc. Thesis ]2], which originated this work. Results obtained with CITER-1D agree very well with the ones found in the the available literature pertaining to WIGL3. Further work on CITER-1D is being pursued, in order to complete the program. (Author) [pt

Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Shim, Hyung Jin

2015-01-01

The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

CRANE: a new scale super-sequence for neutron transport calculations

Energy Technology Data Exchange (ETDEWEB)

Wang, C.; Abdel-Khalik, H.S., E-mail: wang1730@purdue.edu, E-mail: abdelkhalik@purdue.edu [Purdue Univ., School of Nuclear Engineering, West Lafayette, IN (United States); Mertyurek, U., E-mail: umertyurek@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2015-07-01

A new 'super-sequence' called CRANE has been developed to automate the application of reduced order modeling (ROM) to reactor analysis calculations under the SCALE code environment. This new super-sequence is designed to support computationally intensive analyses that require repeated execution of flux solvers with variations in design parameters and nuclear data. This manuscript provides a brief overview of CRANE and demonstrates its applications to representative reactor physics calculations. Specifically, two ROM applications are demonstrated, the intersection subspace-based approach for uncertainty quantification which is intended to reduce the number of uncertainty sources in a conventional uncertainty analysis, and the exact-to-precision generalized perturbation theory methodology intended as a physics-based surrogate model to replace the flux solver, i.e., NEWT. Our overarching goal is to provide a prototypic ROM capability that allows users to further explore and investigate the benefits of using ROM methods in their respective domain and help guide further developments of the methodology and evolution of the tools. (author)

Three-dimensional RAMA fluence methodology benchmarking

International Nuclear Information System (INIS)

Baker, S. P.; Carter, R. G.; Watkins, K. E.; Jones, D. B.

2004-01-01

This paper describes the benchmarking of the RAMA Fluence Methodology software, that has been performed in accordance with U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The RAMA Fluence Methodology has been developed by TransWare Enterprises Inc. through funding provided by the Electric Power Research Inst., Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The purpose of the software is to provide an accurate method for calculating neutron fluence in BWR pressure vessels and internal components. The Methodology incorporates a three-dimensional deterministic transport solution with flexible arbitrary geometry representation of reactor system components, previously available only with Monte Carlo solution techniques. Benchmarking was performed on measurements obtained from three standard benchmark problems which include the Pool Criticality Assembly (PCA), VENUS-3, and H. B. Robinson Unit 2 benchmarks, and on flux wire measurements obtained from two BWR nuclear plants. The calculated to measured (C/M) ratios range from 0.93 to 1.04 demonstrating the accuracy of the RAMA Fluence Methodology in predicting neutron flux, fluence, and dosimetry activation. (authors)

A Quantitative and Systematic Methodology to Investigate Energy Consumption Issues in Multimodal Intercity Transportation Systems

Directory of Open Access Journals (Sweden)

Lili Du

2015-09-01

Full Text Available Energy issues in transportation systems have garnered increasing attention recently. This study proposes a systematic methodology for policy-makers to minimize energy consumption in multimodal intercity transportation systems considering suppliers’ operational constraints and travelers’ mobility requirements. A bi-level optimization model is developed for this purpose and considers the air, rail, private auto, and transit modes. The upper-level model is a mixed integer nonlinear program aiming to minimize energy consumption subject to transportation suppliers’ operational constraints and traffic demand distribution to paths resulting from the lower-level model. The lower-level model is a linear program seeking to maximize the trip utilities of travelers. The interactions between the multimodal transportation suppliers and intercity traffic demand are considered under the goal of minimizing system energy consumption. The proposed bi-level mixed integer model is relaxed and transformed into a mathematical program with complementarity constraints, and solved using a customized branch-and-bound algorithm. Numerical experiments, conducted using multimodal travel options between Lafayette, Indiana and Washington, D.C. reiterate that shifting traffic demand from private cars to the transit and rail modes significantly reduce energy consumption. Moreover, the proposed methodology provides tools to quantitatively analyze system energy consumption and traffic demand distribution among transportation modes under specific policy instruments. The results illustrate the need to systematically incorporate the interactions among traveler preferences, network structure, and supplier operational schemes to provide policy-makers insights for developing traffic demand shift mechanisms to minimize system energy consumption. Hence, the proposed methodology provide policy-makers the capability to analyze energy consumption in the transportation sector by a

Spent Nuclear Fuel Transportation Risk Assessment Methodology for Homeland Security

International Nuclear Information System (INIS)

Teagarden, Grant A.; Canavan, Kenneth T.; Nickell, Robert E.

2006-01-01

In response to increased interest in risk-informed decision making regarding terrorism, EPRI was selected by U.S. DHS and ASME to develop and demonstrate a nuclear sector specific methodology for owner / operators to utilize in performing a Risk Analysis and Management for Critical Asset Protection (RAMCAP) assessment for the transportation of spent nuclear fuel (SNF). The objective is to characterize SNF transportation risk for risk management opportunities and to provide consistent information for DHS decision making. The method uses a characterization of risk as a function of Consequence, Vulnerability, and Threat. Worst reasonable case scenarios characterize risk for a benchmark set of threats and consequence types. A trial application was successfully performed and implementation is underway by one utility. (authors)

Resonance treatment methodology in DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Joo, Han Gyu; Lee, Chung Chan; Chang, Moon Hee

2003-12-01

The typical nuclear design procedure consists of two steps which are the transport lattice calculation for the fuel assembly and the nodal diffusion calculation for the reactor core. DeCART (Deterministic Core Analysis based on Ray Tracing) code has been developed to perform the 3-dimensional whole-core transport calculation removing some of the approximations in the 2-step procedure. This code employs the synthesis of 1- and 2-dimensional characteristics methods in the framework of the 3-dimensional CMFD (Coarse Mesh Finite Difference) formulation. The subgroup method is used for the resonance treatment. HELIOS library is used for the multi-group neutron cross section and the resonance data without any modification. This report includes the methodology of the resonance treatment in DeCART. And this report also includes the Monte Carlo resonance treatment under development for the generation of the resonance integral table and the subgroup data. The interpolation method of the equivalence cross section is reviewed for the efficient resonance transport calculation with thermal-hydraulic feedback, and the new method to consider the temperature distribution explicitly in the subgroup method is also introduced.

Practical methodologies for the calculation of capacity in electricity markets for wind energy

International Nuclear Information System (INIS)

Botero B, Sergio; Giraldo V, Luis Alfonso; Isaza C, Felipe

2008-01-01

Determining the real capacity of the generators in a power market is an essential task in order to estimate the actual system reliability, and to estimate the reward for generators due to their capacity in the firm energy market. In the wind power case, which is an intermittent resource, several methodologies have been proposed to estimate the capacity of a wind power emplacement, not only for planning but also for firm energy remuneration purposes. This paper presents some methodologies that have been proposed or implemented around the world in order to calculate the capacity of this energy resource.

EVALUATION OF ECONOMIC EFFICIENCY PERTAINING TO USAGE OF AUTOMOTIVE TRANSPORT FACILITIES WHILE EXECUTING INTERNATIONAL CARGO TRANSPORTATION

Directory of Open Access Journals (Sweden)

R. B. Ivut

2010-01-01

Full Text Available The paper presents a methodology for evaluation of economic efficiency pertaining to usage of automotive transport facilities while executing international cargo transportation on the basis of average internal norm calculation of automotive operational profitability of a specific model under conditions which are typical for the given market by an average carrier.

Performing three-dimensional neutral particle transport calculations on tera scale computers

International Nuclear Information System (INIS)

Woodward, C.S.; Brown, P.N.; Chang, B.; Dorr, M.R.; Hanebutte, U.R.

1999-01-01

A scalable, parallel code system to perform neutral particle transport calculations in three dimensions is presented. To utilize the hyper-cluster architecture of emerging tera scale computers, the parallel code successfully combines the MPI message passing and paradigms. The code's capabilities are demonstrated by a shielding calculation containing over 14 billion unknowns. This calculation was accomplished on the IBM SP ''ASCI-Blue-Pacific computer located at Lawrence Livermore National Laboratory (LLNL)

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

Optimized shielding calculation to the transport of 131I employed in nuclear medicine

International Nuclear Information System (INIS)

Sahyun, A.; Sordi, G.M.; Rodrigues, D.; Sanches, M.P.; Romero F, C.R.

1996-01-01

The objective of this paper is to present the basis for shielding calculation used in different situations that could occur during the transport of 131 I utilized in nuclear medicine for diagnostic and therapeutic purposes. The aim of these calculation is to optimize the shielding in order to satisfy the transport of radioactive material. These calculations were proposed for estimated activities around 1,85 GBq (50mCi), 3,7 GBq(100mCi) and 7,4 GBq(200mCi), considering the driver of the cargo company and his assistant as the critical group and the general people considered as effect of collective dose. The population density considered in the models is the one related to Sao Paulo city, because the transport is done by the highway across the city and the radioactive material is distributed from west to north and south, where the airports are located. This area ranges a perimeter of 40 km. For the collective dose calculation, it was considered a population dose of less than 1/100 of the annual limit dose for the public. Our main concern is related to the large volume of radioactive material that is transported per week, specially because 1/3 of this material has activities around 3,7 GBq (100mCi). During the calculations, we have figured out that the activities at the moment of transport are nearly 40% greater than the one related to the calibration date. As for the discrepancy of official alpha value of US$10000/man-Sv and the real value for our country of US$3000/man-Sv,a comparative study was performed. (authors). 3 refs., 2 figs., 2 tabs

Chair Report Consultancy Meeting on Nuclear Security Assessment Methodologies (NUSAM) Transport Case Study Working Group

Energy Technology Data Exchange (ETDEWEB)

Shull, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-08-19

The purpose of the consultancy assignment was to (i) apply the NUSAM assessment methods to hypothetical transport security table top exercise (TTX) analyses and (ii) document its results to working materials of NUSAM case study on transport. A number of working group observations, using the results of TTX methodologies, are noted in the report.

Optical photon transport in powdered-phosphor scintillators. Part II. Calculation of single-scattering transport parameters

Energy Technology Data Exchange (ETDEWEB)

Poludniowski, Gavin G. [Joint Department of Physics, Division of Radiotherapy and Imaging, Institute of Cancer Research and Royal Marsden NHS Foundation Trust, Downs Road, Sutton, Surrey SM2 5PT, United Kingdom and Centre for Vision Speech and Signal Processing (CVSSP), Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); Evans, Philip M. [Centre for Vision Speech and Signal Processing (CVSSP), Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom)

2013-04-15

Purpose: Monte Carlo methods based on the Boltzmann transport equation (BTE) have previously been used to model light transport in powdered-phosphor scintillator screens. Physically motivated guesses or, alternatively, the complexities of Mie theory have been used by some authors to provide the necessary inputs of transport parameters. The purpose of Part II of this work is to: (i) validate predictions of modulation transform function (MTF) using the BTE and calculated values of transport parameters, against experimental data published for two Gd{sub 2}O{sub 2}S:Tb screens; (ii) investigate the impact of size-distribution and emission spectrum on Mie predictions of transport parameters; (iii) suggest simpler and novel geometrical optics-based models for these parameters and compare to the predictions of Mie theory. A computer code package called phsphr is made available that allows the MTF predictions for the screens modeled to be reproduced and novel screens to be simulated. Methods: The transport parameters of interest are the scattering efficiency (Q{sub sct}), absorption efficiency (Q{sub abs}), and the scatter anisotropy (g). Calculations of these parameters are made using the analytic method of Mie theory, for spherical grains of radii 0.1-5.0 {mu}m. The sensitivity of the transport parameters to emission wavelength is investigated using an emission spectrum representative of that of Gd{sub 2}O{sub 2}S:Tb. The impact of a grain-size distribution in the screen on the parameters is investigated using a Gaussian size-distribution ({sigma}= 1%, 5%, or 10% of mean radius). Two simple and novel alternative models to Mie theory are suggested: a geometrical optics and diffraction model (GODM) and an extension of this (GODM+). Comparisons to measured MTF are made for two commercial screens: Lanex Fast Back and Lanex Fast Front (Eastman Kodak Company, Inc.). Results: The Mie theory predictions of transport parameters were shown to be highly sensitive to both grain size

Time-dependent Flow and Transport Calculations for Project Opalinus Clay (Entsorgungsnachweis)

International Nuclear Information System (INIS)

Kosakowski, G.

2004-07-01

This report describes two specific assessment cases used in the safety assessment for a proposed deep geological repository for spent fuel, high level waste and long-lived intermediate-level waste, sited in the Opalinus Clay of the Zuercher Weinland in northern Switzerland (Project Entsorgungsnachweis, NAG RA, 2002d). In this study the influence of time dependent flow processes on the radionuclide transport in the geosphere is investigated. In the Opalinus Clay diffusion dominates the transport of radionuclides, but processes exist that can locally increase the importance of the advective transport for some time. Two important cases were investigated: (1) glaciation-induced flow due to an additional overburden in the form of an ice shield of up to 400 m thickness and (2) fluid flow driven by tunnel convergence. For the calculations the code FRAC3DVS (Therrien and Sudicky, 1996) was used. FRAC3DVS solves the three-dimensional flow and transport equation in porous and fractured media. For the case of glaciation-induced flow (1) a two-dimensional reference model without glaciations was calculated. During the glaciations the geosphere release-rates are up to a factor of about 1.7 higher compared to the reference model. The influence of glaciations on the transport of cations or neutral species is less than for anions, since the importance of the advective transport for anions is higher due to the lower accessible porosity for anions. The increase in the release rates during glaciations is lower for sorbing compared to non-sorbing radionuclides. The influence of the tunnel convergence (2) on the transport of radionuclides in the geosphere is very small. Due to the higher source term the geosphere release rates are slightly higher if tunnel convergence is considered. In addition to the two assessment cases this report investigates the applicability of the one-dimensional approximation for modelling transport through the Opalinus Clay. For the reference case of the safety

Axial SPN and radial MOC coupled whole core transport calculation

International Nuclear Information System (INIS)

Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan; Zee, Sung-Quun; Joo, Han-Gyu

2007-01-01

The Simplified P N (SP N ) method is applied to the axial solution of the two-dimensional (2-D) method of characteristics (MOC) solution based whole core transport calculation. A sub-plane scheme and the nodal expansion method (NEM) are employed for the solution of the one-dimensional (1-D) SP N equations involving a radial transverse leakage. The SP N solver replaces the axial diffusion solver of the DeCART direct whole core transport code to provide more accurate, transport theory based axial solutions. In the sub-plane scheme, the radial equivalent homogenization parameters generated by the local MOC for a thick plane are assigned to the multiple finer planes in the subsequent global three-dimensional (3-D) coarse mesh finite difference (CMFD) calculation in which the NEM is employed for the axial solution. The sub-plane scheme induces a much less nodal error while having little impact on the axial leakage representation of the radial MOC calculation. The performance of the sub-plane scheme and SP N nodal transport solver is examined by solving a set of demonstrative problems and the C5G7MOX 3-D extension benchmark problems. It is shown in the demonstrative problems that the nodal error reaching upto 1,400 pcm in a rodded case is reduced to 10 pcm by introducing 10 sub-planes per MOC plane and the transport error is reduced from about 150 pcm to 10 pcm by using SP 3 . Also it is observed, in the C5G7MOX rodded configuration B problem, that the eigenvalues and pin power errors of 180 pcm and 2.2% of the 10 sub-planes diffusion case are reduced to 40 pcm and 1.4%, respectively, for SP 3 with only about a 15% increase in the computing time. It is shown that the SP 5 case gives very similar results to the SP 3 case. (author)

Iterative resonance self-shielding methods using resonance integral table in heterogeneous transport lattice calculations

International Nuclear Information System (INIS)

Hong, Ser Gi; Kim, Kang-Seog

2011-01-01

This paper describes the iteration methods using resonance integral tables to estimate the effective resonance cross sections in heterogeneous transport lattice calculations. Basically, these methods have been devised to reduce an effort to convert resonance integral table into subgroup data to be used in the physical subgroup method. Since these methods do not use subgroup data but only use resonance integral tables directly, these methods do not include an error in converting resonance integral into subgroup data. The effective resonance cross sections are estimated iteratively for each resonance nuclide through the heterogeneous fixed source calculations for the whole problem domain to obtain the background cross sections. These methods have been implemented in the transport lattice code KARMA which uses the method of characteristics (MOC) to solve the transport equation. The computational results show that these iteration methods are quite promising in the practical transport lattice calculations.

Scaling up methodology for CO2 emissions in ICT applications in traffic and transport in Europe

NARCIS (Netherlands)

Mans, D.; Jonkers, E.; Giannelos, I.; Palanciuc, D.

2013-01-01

The Amitran project aims to define a reference methodology for evaluating the effects of ICT measures in trafäc and transport on energy efficiency and consequently CO2 emissions. This methodology can be used as a reference by future projects and will address different modes for both passenger and

Evaluation of doses and risks from different decontamination and decommissioning strategies using the PRESTO-II methodology

International Nuclear Information System (INIS)

Fields, D.E.

1986-01-01

The PRESTO-II methodology may be applied to evaluate doses and health risks from a variety of decontamination and decommissioning activities. This methodology has been implemented in the form of a computer code that has been applied to several sites, and that has been extensively documented. Radionuclide inventories are specified as separate contamination sources either present on the ground surface, covered by non-radioactive soils but lying above the water table, suspended in the atmosphere, or dissolved in surface waters. Hydrologic transport mechanisms considered in the PRESTO-II methodology include chemical exchange, ponding and overflow, surface water transport, groundwater transport, and pumping contaminated groundwater from wells. Varied scenarios of water usage are treated. Atmospheric inputs are based on both resuspension factor and resuspension rate approaches, with inhalation and immersion doses based on a Gaussian plume transport calculation. Site activities that are considered include land clearing, farming, and residing on the site. Exposure and dose calculations are derived from the US Nuclear Regulatory Commission Reg. Guide 1.109 approach, while risk calculations use a life-table approach developed for the US Environmental Protection Agency (EPA). Internal dose conversion factors are taken from ICRP 26 and 30, while risk conversion factors are values suggested by EPA. 19 refs., 2 figs., 1 tab

Survey of shielding calculation parameters in radiotherapy rooms used in the country and its impact in the existing calculation methodologies

International Nuclear Information System (INIS)

Japiassu, Fernando Parois

2013-01-01

When designing radiotherapy treatment rooms, the dimensions of barriers are established on the basis of American calculation methodologies specifically; NCRP Report N° 49, NCRP Report N° 51, and more recently, NCRP Report N° 151. Such barrier calculations are based on parameters reflecting predictions of treatments to be performed within the room; which, in tum, reftect a specific reality found in a country. There exists, however, a variety of modern radiotherapy techniques, such as Intensity Modulated Radiation Therapy (IMRT); Total Body Irradiation (TBl) and radiosurgery (SRS); where patierits are treated in a much different way than during more conventional treatrnents, which are not taken into account the traditional shielding calculation methodology. This may lead to a faulty design of treattnent rooms. In order to establish a comparison between the methodology used to calculate shielding design and the reality of treatments performed in Brazil, two radiotherapy facilitie were selected, both of them offering traditional and modern treatment techniqued as described above. Data in relation with reatments perfotmed over a period of six (6)months of operations in both institutions were collected. Based on tlis informaton, a new set of realistic parameters required for shielding design was estãblished, whicb in turn allowed for a nwe caculation of barrier thickness for both facilities. The barrier thickness resultaing from this calculation was then compared with the barrier thickness propose as part of the original shielding design, approved by the regulatory authority. First, concerning the public facility, the thickness of all primary barriers proposed in the shielding design was actually larger than the thickness resulting from calculations based on realistic parameters. Second, concerning the private facility, the new data show that the thickness of three out of the four primary barriers described in the project is larger than the thickness oresulting from

Methodology of external exposure calculation for reuse of conditional released materials from decommissioning - 59138

International Nuclear Information System (INIS)

Ondra, Frantisek; Vasko, Marek; Necas, Vladimir

2012-01-01

The article presents methodology of external exposure calculation for reuse of conditional released materials from decommissioning using VISIPLAN 3D ALARA planning tool. Production of rails has been used as an example application of proposed methodology within the CONRELMAT project. The article presents a methodology for determination of radiological, material, organizational and other conditions for conditionally released materials reuse to ensure that workers and public exposure does not breach the exposure limits during scenario's life cycle (preparation, construction and operation of scenario). The methodology comprises a proposal of following conditions in the view of workers and public exposure: - radionuclide limit concentration of conditionally released materials for specific scenarios and nuclide vectors, - specific deployment of conditionally released materials eventually shielding materials, workers and public during the scenario's life cycle, - organizational measures concerning time of workers or public stay in the vicinity on conditionally released materials for individual performed scenarios and nuclide vectors. The above mentioned steps of proposed methodology have been applied within the CONRELMAT project. Exposure evaluation of workers for rail production is introduced in the article as an example of this application. Exposure calculation using VISIPLAN 3D ALARA planning tool was done within several models. The most exposed profession for scenario was identified. On the basis of this result, an increase of radionuclide concentration in conditional released material was proposed more than two times to 681 Bq/kg without no additional safety or organizational measures being applied. After application of proposed safety and organizational measures (additional shielding, geometry changes and limitation of work duration) it is possible to increase concentration of radionuclide in conditional released material more than ten times to 3092 Bq/kg. Storage

Calculations of Neutron Flux Distributions by Means of Integral Transport Methods

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1967-05-15

Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.

Design of a transport calculation system for logging sondes simulation

International Nuclear Information System (INIS)

Marquez Damian, Jose Ignacio

2005-01-01

Analysis of available resources in earth crust is performed by different techniques, one of them is neutron logging. Design of sondes that are used to make such logging is supported by laboratory experiments as well as by numerical calculations.This work presents several calculation schemes, designed to simplify the task of whom has to planify such experiments or optimize parameters of this kind of sondes.These schemes use transport calculation codes, especially DaRT, TORT and MCNP, and cross section processing modules from SCALE system.Additionally a system for DaRT and TORT data postprocessing using OpenDX is presented.It allows scalar flux spatial distribution analysis, as wells as cross section condensation and reaction rates calculation

A methodology for on-line calculation of temperature and thermal stress under non-linear boundary conditions

International Nuclear Information System (INIS)

Botto, D.; Zucca, S.; Gola, M.M.

2003-01-01

In the literature many works have been written dealing with the task of on-line calculation of temperature and thermal stress for machine components and structures, in order to evaluate fatigue damage accumulation and estimate residual life. One of the most widespread methodologies is the Green's function technique (GFT), by which machine parameters such as fluid temperatures, pressures and flow rates are converted into metal temperature transients and thermal stresses. However, since the GFT is based upon the linear superposition principle, it cannot be directly used in the case of varying heat transfer coefficients. In the present work, a different methodology is proposed, based upon CMS for temperature transient calculation and upon the GFT for the related thermal stress evaluation. This new approach allows variable heat transfer coefficients to be accounted for. The methodology is applied for two different case studies, taken from the literature: a thick pipe and a nozzle connected to a spherical head, both subjected to multiple convective boundary conditions

AUTOMATION OF CALCULATION ALGORITHMS FOR EFFICIENCY ESTIMATION OF TRANSPORT INFRASTRUCTURE DEVELOPMENT

Directory of Open Access Journals (Sweden)

Sergey Kharitonov

2015-06-01

Full Text Available Optimum transport infrastructure usage is an important aspect of the development of the national economy of the Russian Federation. Thus, development of instruments for assessing the efficiency of infrastructure is impossible without constant monitoring of a number of significant indicators. This work is devoted to the selection of indicators and the method of their calculation in relation to the transport subsystem as airport infrastructure. The work also reflects aspects of the evaluation of the possibilities of algorithmic computational mechanisms to improve the tools of public administration transport subsystems.

Methodology to Calculate the Costs of a Floating Offshore Renewable Energy Farm

Directory of Open Access Journals (Sweden)

Laura Castro-Santos

2016-04-01

Full Text Available This paper establishes a general methodology to calculate the life-cycle cost of floating offshore renewable energy devices, applying it to wave energy and wind energy devices. It is accounts for the contributions of the six main phases of their life-cycle: concept definition, design and development, manufacturing, installation, exploitation and dismantling, the costs of which have been defined. Moreover, the energy produced is also taken into account to calculate the Levelized Cost of Energy of a floating offshore renewable energy farm. The methodology proposed has been applied to two renewable energy devices: a floating offshore wave energy device and a floating offshore wind energy device. Two locations have been considered: Aguçadoura and São Pedro de Moel, both in Portugal. Results indicate that the most important cost in terms of the life-cycle of a floating offshore renewable energy farm is the exploitation cost, followed by the manufacturing and the installation cost. In addition, the best area in terms of costs is the same independently of the type of floating offshore renewable energy considered: Aguçadoura. However, the results in terms of Levelized Cost of Energy are different: Aguçadoura is better when considering wave energy technology and the São Pedro de Moel region is the best option when considering floating wind energy technology. The method proposed aims to give a direct approach to calculate the main life-cycle cost of a floating offshore renewable energy farm. It helps to assess its feasibility and evaluating the relevant characteristics that influence it the most.

Methodology of strength calculation under alternating stresses using the diagram of limiting amplitudes

Science.gov (United States)

Konovodov, V. V.; Valentov, A. V.; Kukhar, I. S.; Retyunskiy, O. Yu; Baraksanov, A. S.

2016-08-01

The work proposes the algorithm to calculate strength under alternating stresses using the developed methodology of building the diagram of limiting stresses. The overall safety factor is defined by the suggested formula. Strength calculations of components working under alternating stresses in the great majority of cases are conducted as the checking ones. It is primarily explained by the fact that the overall fatigue strength reduction factor (Kσg or Kτg) can only be chosen approximately during the component design as the engineer at this stage of work has just the approximate idea on the component size and shape.

International report to validate criticality safety calculations for fissile material transport

International Nuclear Information System (INIS)

Whitesides, G.E.

1984-01-01

During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

A new methodology for modeling of direct landslide costs for transportation infrastructures

Science.gov (United States)

Klose, Martin; Terhorst, Birgit

2014-05-01

The world's transportation infrastructure is at risk of landslides in many areas across the globe. A safe and affordable operation of traffic routes are the two main criteria for transportation planning in landslide-prone areas. The right balancing of these often conflicting priorities requires, amongst others, profound knowledge of the direct costs of landslide damage. These costs include capital investments for landslide repair and mitigation as well as operational expenditures for first response and maintenance works. This contribution presents a new methodology for ex post assessment of direct landslide costs for transportation infrastructures. The methodology includes tools to compile, model, and extrapolate landslide losses on different spatial scales over time. A landslide susceptibility model enables regional cost extrapolation by means of a cost figure obtained from local cost compilation for representative case study areas. On local level, cost survey is closely linked with cost modeling, a toolset for cost estimation based on landslide databases. Cost modeling uses Landslide Disaster Management Process Models (LDMMs) and cost modules to simulate and monetize cost factors for certain types of landslide damage. The landslide susceptibility model provides a regional exposure index and updates the cost figure to a cost index which describes the costs per km of traffic route at risk of landslides. Both indexes enable the regionalization of local landslide losses. The methodology is applied and tested in a cost assessment for highways in the Lower Saxon Uplands, NW Germany, in the period 1980 to 2010. The basis of this research is a regional subset of a landslide database for the Federal Republic of Germany. In the 7,000 km² large Lower Saxon Uplands, 77 km of highway are located in potential landslide hazard area. Annual average costs of 52k per km of highway at risk of landslides are identified as cost index for a local case study area in this region. The

Understanding how transport choices are affected by the environment and health: views expressed in a study on the use of carbon calculators.

Science.gov (United States)

Chatterton, T J; Coulter, A; Musselwhite, C; Lyons, G; Clegg, S

2009-01-01

To examine the influence that the provision of environmental information might be able to make on personal travel behaviour through analysis of the views of members of the public expressed in a study for the UK Department for Transport on attitudes towards carbon calculator tools. A three-stage qualitative survey taking an ideographic approach to analysing public attitudes to the use of carbon calculator tools in relation to making transport decisions. Interviews and discussion groups with stakeholders, non-users and users providing extensive data that were analysed using the British Market Research Bureau's matrix mapping methodology. Despite considerable awareness of climate change as an issue, personal carbon emissions were not found to have much influence on personal transport choice, which could be seen as being dominated by issues of cost (both in time and money), comfort and convenience. The spatial and temporal dislocation of the cause and effects of climate change make it difficult to link the impacts of personal travel behaviour with specific activities. If environmental- and health-based information is to be provided as a lever to change travel behaviour, it may be necessary to provide information on issues such as local air pollution and personal health impacts in order to link wider benefits with a travel user's self-interest.

Resonance Self-Shielding Methodologies in SCALE 6

International Nuclear Information System (INIS)

Williams, Mark L.

2011-01-01

SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.

76 FR 12985 - Request for Comments on Trend Factor Methodology Used in the Calculation of Fair Market Rents

Science.gov (United States)

2011-03-09

... Trend Factor Methodology Used in the Calculation of Fair Market Rents AGENCY: Office of the Assistant... used to calculate the trend factor component of the Fair Market Rent estimates. SUMMARY: Section 8(c)(1... comment regarding the manner in which HUD calculates the trend factor used in the Fair Market Rent (FMR...

FPGA hardware acceleration for high performance neutron transport computation based on agent methodology - 318

International Nuclear Information System (INIS)

Shanjie, Xiao; Tatjana, Jevremovic

2010-01-01

The accurate, detailed and 3D neutron transport analysis for Gen-IV reactors is still time-consuming regardless of advanced computational hardware available in developed countries. This paper introduces a new concept in addressing the computational time while persevering the detailed and accurate modeling; a specifically designed FPGA co-processor accelerates robust AGENT methodology for complex reactor geometries. For the first time this approach is applied to accelerate the neutronics analysis. The AGENT methodology solves neutron transport equation using the method of characteristics. The AGENT methodology performance was carefully analyzed before the hardware design based on the FPGA co-processor was adopted. The most time-consuming kernel part is then transplanted into the FPGA co-processor. The FPGA co-processor is designed with data flow-driven non von-Neumann architecture and has much higher efficiency than the conventional computer architecture. Details of the FPGA co-processor design are introduced and the design is benchmarked using two different examples. The advanced chip architecture helps the FPGA co-processor obtaining more than 20 times speed up with its working frequency much lower than the CPU frequency. (authors)

TRING: a computer program for calculating radionuclide transport in groundwater

International Nuclear Information System (INIS)

Maul, P.R.

1984-12-01

The computer program TRING is described which enables the transport of radionuclides in groundwater to be calculated for use in long term radiological assessments using methods described previously. Examples of the areas of application of the program are activity transport in groundwater associated with accidental spillage or leakage of activity, the shutdown of reactors subject to delayed decommissioning, shallow land burial of intermediate level waste and geologic disposal of high level waste. Some examples of the use of the program are given, together with full details to enable users to run the program. (author)

Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

Energy Technology Data Exchange (ETDEWEB)

Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)

2012-07-01

In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

International Nuclear Information System (INIS)

Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.

2012-01-01

In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

CERN Document Server

2002-01-01

This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

How can activity-based costing methodology be performed as a powerful tool to calculate costs and secure appropriate patient care?

Science.gov (United States)

Lin, Blossom Yen-Ju; Chao, Te-Hsin; Yao, Yuh; Tu, Shu-Min; Wu, Chun-Ching; Chern, Jin-Yuan; Chao, Shiu-Hsiung; Shaw, Keh-Yuong

2007-04-01

Previous studies have shown the advantages of using activity-based costing (ABC) methodology in the health care industry. The potential values of ABC methodology in health care are derived from the more accurate cost calculation compared to the traditional step-down costing, and the potentials to evaluate quality or effectiveness of health care based on health care activities. This project used ABC methodology to profile the cost structure of inpatients with surgical procedures at the Department of Colorectal Surgery in a public teaching hospital, and to identify the missing or inappropriate clinical procedures. We found that ABC methodology was able to accurately calculate costs and to identify several missing pre- and post-surgical nursing education activities in the course of treatment.

REVIEW OF METHODOLOGIES FOR COSTS CALCULATING OF RUMINANTS IN SLOVAKIA

Directory of Open Access Journals (Sweden)

Zuzana KRUPOVÁ

2012-09-01

Full Text Available The objective of this work was to synthesise and analyse the methodologies and the biological aspects of the costs calculation in ruminants in Slovakia. According to literature, the account classification of cost items is most often considered for construction of costing formula. The costs are mostly divided into fixed (costs independent from volume of herd’s production and variable ones (costs connected with improvement of breeding conditions. Cost for feeds and beddings, labour costs, other direct costs and depreciations were found as the most important cost items in ruminants. It can be assumed that including the depreciations into costs of the basic herd takes into consideration the real costs simultaneously invested into raising of young animals in the given period. Costs are calculated for the unit of the main and by-products and their classification is influenced mainly by the type of livestock and production system. In dairy cows is usually milk defined as the main product, and by- products are live born calf and manure. The base calculation unit is kilogram of milk (basic herd of cows and kilogram of gain and kilogram of live weight (young breeding cattle. In suckler cows is a live-born calf the main product and manure is the by-product. The costs are mostly calculated per suckler cow, live-born calf and per kilogram of live weight of weaned calf. Similar division of products into main and by-products is also in cost calculation for sheep categories. The difference is that clotted cheese is also considered as the main product of basic herd in dairy sheep and greasy wool as the by-products in all categories. Definition of the base calculation units in sheep categories followed the mentioned classification. The value of a by-product in cattle and sheep is usually set according to its quantity and intra- plant price of the by-product. In the calculation of the costs for sheep and cattle the “structural ewe” and “structural cow�

A methodology for assessing social considerations in transport of low and intermediate level radioactive waste

International Nuclear Information System (INIS)

Allsop, R.E.; Banister, D.J.; Holden, D.J.; Bird, J.; Downe, H.E.

1986-05-01

A methodology is proposed for taking into account non-radiological social aspects of the transport of low and intermediate level radioactive waste when considering the location of disposal facilities and the transport of waste to such facilities from the sites where it arises. As part of a data acquisition programme, an attitudinal survey of a sample of people unconnected with any suggested site or transport route is proposed in order to estimate levels of concern felt by people of different kinds about waste transport. Probabilities of accident occurrence during transport by road and rail are also discussed, and the limited extent of quantified information about consequences of accidents is reviewed. The scope for malicious interference with consignments of waste in transit is considered. (author)

Improved method for calculating neoclassical transport coefficients in the banana regime

Energy Technology Data Exchange (ETDEWEB)

Taguchi, M., E-mail: taguchi.masayoshi@nihon-u.ac.jp [College of Industrial Technology, Nihon University, Narashino 275-8576 (Japan)

2014-05-15

The conventional neoclassical moment method in the banana regime is improved by increasing the accuracy of approximation to the linearized Fokker-Planck collision operator. This improved method is formulated for a multiple ion plasma in general tokamak equilibria. The explicit computation in a model magnetic field shows that the neoclassical transport coefficients can be accurately calculated in the full range of aspect ratio by the improved method. The some neoclassical transport coefficients for the intermediate aspect ratio are found to appreciably deviate from those obtained by the conventional moment method. The differences between the transport coefficients with these two methods are up to about 20%.

Safety, mobility and comfort assessment methodologies of intelligent transport systems for vulnerable road users

NARCIS (Netherlands)

Malone, K.; Silla, A.; Johanssen, C.; Bell, D.

2017-01-01

Introduction: This paper describes the modification and development of methodologies to assess the impacts of Intelligent Transport Systems (ITS) applications for Vulnerable Road users (VRUs) in the domains of safety, mobility and comfort. This effort was carried out in the context of the VRUITS

Evaluation of radiation shielding performance in sea transport of radioactive material by using simple calculation method

International Nuclear Information System (INIS)

Odano, N.; Ohnishi, S.; Sawamura, H.; Tanaka, Y.; Nishimura, K.

2004-01-01

A modified code system based on the point kernel method was developed to use in evaluation of shielding performance for maritime transport of radioactive material. For evaluation of shielding performance accurately in the case of accident, it is required to preciously model the structure of transport casks and shipping vessel, and source term. To achieve accurate modelling of the geometry and source term condition, we aimed to develop the code system by using equivalent information regarding structure and source term used in the Monte Carlo calculation code, MCNP. Therefore, adding an option to use point kernel method to the existing Monte Carlo code, MCNP4C, the code system was developed. To verify the developed code system, dose rate distribution in an exclusive shipping vessel to transport the low level radioactive wastes were calculated by the developed code and the calculated results were compared with measurements and Monte Carlo calculations. It was confirmed that the developed simple calculation method can obtain calculation results very quickly with enough accuracy comparing with the Monte Carlo calculation code MCNP4C

Considerations of beta and electron transport in internal dose calculations. Progress report

Energy Technology Data Exchange (ETDEWEB)

Bolch, W.E.

1994-11-01

The goal of this particular task is to consider, for the first time, the explicit transport of beta particles and photon-generated electrons in the series of six phantoms developed by Cristy and Eckerman (1987) at the Oak Ridge National Laboratory. In their report, ORNL/TM-8381, specific absorbed fractions of energy are reported for phantoms representing the newborn (3.4 kg), the one-year-old (9.8 kg), the five-year-old (19 kg), the ten-year-old (32 kg), the fifteen-year-old/adult female (55-58 kg), and the adult male (70 kg). Radiation transport calculations were performed with the Monte Carlo code ALGAMP which allows photon transport only. In subsequent calculations of radionuclide S values as is done in the MIRDOSE2 computer program, electron absorbed fractions are thus considered to be either unity or zero depending upon whether the source region does or does not equal the target region, respectively.

Beam transport calculations for BARC-TIFR 14UD pelletron

International Nuclear Information System (INIS)

Prasad, K.G.

1993-01-01

The 14UD pelletron tandem accelerator installed at Tata Institute of Fundamental Research (TIFR) as a joint BARC-TIFR project, is supplied by National Electrostatic Corporation (NEC), U.S.A. To optimise the parameters of various elements along the beam path, it is essential to work out the beam optics of the entire system. There are various computer codes in use for such calculations. All these codes, except the detailed ray tracing programs, use matrix formulation. Thus each ion optical element is characterised in terms of a transport matrix, whose elements are assumed to be independent of particle trajectory. We have performed only the first order calculations, meaning thereby that no aberrations are included. Further, all calculations are carried out assuming ideal conditions like axial beam injection, perfectly aligned beam line elements, etc. The main code that has been employed in our calculations is based on the one at the Australian National University, Canberra, suitably modified for use with CYBER 170/730 computer at TIFR. However, codes at NEC and Stony Brook were also used for the checking the results. The results of calculations are given and discussed. (author). 2 figs

Program for calculating multi-component high-intense ion beam transport

International Nuclear Information System (INIS)

Kazarinov, N.Yu.; Prejzendorf, V.A.

1985-01-01

The CANAL program for calculating transport of high-intense beams containing ions with different charges in a channel consisting of dipole magnets and quadrupole lenses is described. The equations determined by the method of distribution function momenta and describing coordinate variations of the local mass centres and r.m.s. transverse sizes of beams with different charges form the basis of the calculation. The program is adapted for the CDC-6500 and SM-4 computers. The program functioning is organized in the interactive mode permitting to vary the parameters of any channel element and quickly choose the optimum version in the course of calculation. The calculation time for the CDC-6500 computer for the 30-40 m channel at the integration step of 1 cm is about 1 min. The program is used for calculating the channel for the uranium ion beam injection from the collective accelerator into the heavy-ion synchrotron

Space-Time Dependent Transport, Activation, and Dose Rates for Radioactivated Fluids.

Science.gov (United States)

Gavazza, Sergio

Two methods are developed to calculate the space - and time-dependent mass transport of radionuclides, their production and decay, and the associated dose rates generated from the radioactivated fluids flowing through pipes. The work couples space- and time-dependent phenomena, treated as only space- or time-dependent in the open literature. The transport and activation methodology (TAM) is used to numerically calculate space- and time-dependent transport and activation of radionuclides in fluids flowing through pipes exposed to radiation fields, and volumetric radioactive sources created by radionuclide motions. The computer program Radionuclide Activation and Transport in Pipe (RNATPA1) performs the numerical calculations required in TAM. The gamma ray dose methodology (GAM) is used to numerically calculate space- and time-dependent gamma ray dose equivalent rates from the volumetric radioactive sources determined by TAM. The computer program Gamma Ray Dose Equivalent Rate (GRDOSER) performs the numerical calculations required in GAM. The scope of conditions considered by TAM and GAM herein include (a) laminar flow in straight pipe, (b)recirculating flow schemes, (c) time-independent fluid velocity distributions, (d) space-dependent monoenergetic neutron flux distribution, (e) space- and time-dependent activation process of a single parent nuclide and transport and decay of a single daughter radionuclide, and (f) assessment of space- and time-dependent gamma ray dose rates, outside the pipe, generated by the space- and time-dependent source term distributions inside of it. The methodologies, however, can be easily extended to include all the situations of interest for solving the phenomena addressed in this dissertation. A comparison is made from results obtained by the described calculational procedures with analytical expressions. The physics of the problems addressed by the new technique and the increased accuracy versus non -space and time-dependent methods

Review on the NEI Methodology of Debris Transport Analysis in Sump Blockage Issue for APR1400

International Nuclear Information System (INIS)

Kim, Jong Uk; Lee, Jeong Ik; Hong, Soon Joon; Lee, Byung Chul; Bang, Young Seok

2007-01-01

Since USNRC (United State Nuclear Regulatory Committee) initially addressed post-accident sump performance under Unresolved Safety Issue USI A-43, sump blockage issue has gone through GSI-191, Regulation Guide 1.82, Rev. 3 (RG. 1.82 Rev.3), and generic Letter 2004-02 for PWRs (Pressurized Water Reactors). As a response of these USNRC's activities, NEI 04-07 was issued in order to evaluate the post-accident performance of a plant's recirculation sump. The baseline methodology of NEI 04-07 is composed of break selection, debris generation, latent debris, debris transport, and head loss. In analytical refinement of NEI 04-07, computational fluid dynamic (CFD) is suggested for the evaluation of debris transport in emergency core cooling (ECC) recirculation mode as guided by RG. 1.82 Rev.3. In Korea nuclear industry also keeps step with international activities of this safety issue, with Kori 1 plant as a pioneering edge. Korean nuclear industry has been also pursuing development of an advanced PWR of APR1400, which incorporates several improved safety features. One of the key features, considering sump blockage issue, is the adoption of IRWST (In-containment Refueling Water Storage Tank). This device, as the acronym implies, changes the emergency core cooling water injection pattern. This fact makes us to review the applicability of NEI 04-07's methodology. In this paper we discuss the applicability of NEI 04- 07's methodology, and more over, new methodology is proposed. And finally the preliminary debris transport is analyzed

Risks of transport of radioactive materials on the road; some exploring calculations performed with the INTERTRAN-model

International Nuclear Information System (INIS)

1987-04-01

Under the auspices of the IAEA a computercode, named INTERTRAN, has been developed in order to calculate the risks of the transport of radioactive materials. This code has to be tested nearer. For the Dutch situation a number of calculations has been performed of more or less realistic cases in which four transport streams have been investigated. Two transport routes are chosen. The risks thus obtained are compared quantitatively with the risks of LPG-transports. 4 refs.; 9 figs

Ab Initio Calculations of Transport Properties of Vanadium Oxides

Science.gov (United States)

Lamsal, Chiranjivi; Ravindra, N. M.

2018-04-01

The temperature-dependent transport properties of vanadium oxides have been studied near the Fermi energy using the Kohn-Sham band structure approach combined with Boltzmann transport equations. V2O5 exhibits significant thermoelectric properties, which can be attributed to its layered structure and stability. Highly anisotropic electrical conduction in V2O5 is clearly manifested in the calculations. Due to specific details of the band structure and anisotropic electron-phonon interactions, maxima and crossovers are also seen in the temperature-dependent Seebeck coefficient of V2O5. During the phase transition of VO2, the Seebeck coefficient changes by 18.9 µV/K, which is close to (within 10% of) the observed discontinuity of 17.3 µV/K.

A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

International Nuclear Information System (INIS)

Rossini, M.R.

1992-01-01

An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

Simulation and Optimization Methodologies for Military Transportation Network Routing and Scheduling and for Military Medical Services

National Research Council Canada - National Science Library

Rodin, Ervin Y

2005-01-01

The purpose of this present research was to develop a generic model and methodology for analyzing and optimizing large-scale air transportation networks including both their routing and their scheduling...

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

Development of new methodology for dose calculation in photographic dosimetry

International Nuclear Information System (INIS)

Daltro, T.F.L.

1994-01-01

A new methodology for equivalent dose calculations has been developed at IPEN-CNEN/SP to be applied at the Photographic Dosimetry Laboratory using artificial intelligence techniques by means of neutral network. The research was orientated towards the optimization of the whole set of parameters involves in the film processing going from the irradiation in order to obtain the calibration curve up to the optical density readings. The learning of the neutral network was performed by taking the readings of optical density from calibration curve as input and the effective energy and equivalent dose as output. The obtained results in the intercomparison show an excellent agreement with the actual values of dose and energy given by the National Metrology Laboratory of Ionizing Radiation. (author)

Development of new methodology for dose calculation in photographic dosimetry

International Nuclear Information System (INIS)

Daltro, T.F.L.; Campos, L.L.

1994-01-01

A new methodology for equivalent dose calculation has been developed at IPEN-CNEN/SP to be applied at the Photographic Dosimetry Laboratory using artificial intelligence techniques by means of neural network. The research was oriented towards the optimization of the whole set of parameters involved in the film processing going from the irradiation in order to obtain the calibration curve up to the optical density readings. The learning of the neural network was performed by taking readings of optical density from calibration curve as input and the effective energy and equivalent dose as output. The obtained results in the intercomparison show an excellent agreement with the actual values of dose and energy given by the National Metrology Laboratory of Ionizing Radiation

Parallel MCNP Monte Carlo transport calculations with MPI

International Nuclear Information System (INIS)

Wagner, J.C.; Haghighat, A.

1996-01-01

The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected

Simplified calculation method for radiation dose under normal condition of transport

International Nuclear Information System (INIS)

Watabe, N.; Ozaki, S.; Sato, K.; Sugahara, A.

1993-01-01

In order to estimate radiation dose during transportation of radioactive materials, the following computer codes are available: RADTRAN, INTERTRAN, J-TRAN. Because these codes consist of functions for estimating doses not only under normal conditions but also in the case of accidents, when nuclei may leak and spread into the environment by air diffusion, the user needs to have special knowledge and experience. In this presentation, we describe how, with a view to preparing a method by which a person in charge of transportation can calculate doses in normal conditions, the main parameters upon which the value of doses depends were extracted and the dose for a unit of transportation was estimated. (J.P.N.)

A proposed methodology for the calculation of direct consumption of fossil fuels and electricity for livestock breeding, and its application to Cyprus

International Nuclear Information System (INIS)

Kythreotou, Nicoletta; Florides, Georgios; Tassou, Savvas A.

2012-01-01

On-farm energy consumption is becoming increasingly important in the context of rising energy costs and concerns over greenhouse gas emissions. For farmers throughout the world, energy inputs represent a major and rapidly increasing cost. In many countries such as Cyprus, however, there is lack of systematic research on energy use in agriculture, which hinders benchmarking end evaluation of approaches and investment decisions for energy improvement. This study established a methodology for the estimation of the direct consumption of fossil fuels and electricity for livestock breeding, excluding transport, for locations where full data sets are not available. This methodology was then used to estimate fossil fuel and electricity consumption for livestock breeding in Cyprus. For 2008, this energy was found to be equivalent to 40.3 GWh that corresponds to 8% of the energy used in agriculture. Differences between the energy consumption per animal in Cyprus and other countries was found to be mainly due to differences in climatic conditions and technologies used in the farms. -- Highlights: ► A methodology to calculate energy consumption in farming applied to Cyprus. ► Annual consumption per animal was estimated to be 565 kWh/cow, 537 kWh/sow and 0.677 kWh/chicken. ► Direct energy consumption in livestock breeding is estimated at 40.3 GWh in 2008.

Thermal transport across metal silicide-silicon interfaces: First-principles calculations and Green's function transport simulations

Science.gov (United States)

Sadasivam, Sridhar; Ye, Ning; Feser, Joseph P.; Charles, James; Miao, Kai; Kubis, Tillmann; Fisher, Timothy S.

2017-02-01

Heat transfer across metal-semiconductor interfaces involves multiple fundamental transport mechanisms such as elastic and inelastic phonon scattering, and electron-phonon coupling within the metal and across the interface. The relative contributions of these different transport mechanisms to the interface conductance remains unclear in the current literature. In this work, we use a combination of first-principles calculations under the density functional theory framework and heat transport simulations using the atomistic Green's function (AGF) method to quantitatively predict the contribution of the different scattering mechanisms to the thermal interface conductance of epitaxial CoSi2-Si interfaces. An important development in the present work is the direct computation of interfacial bonding from density functional perturbation theory (DFPT) and hence the avoidance of commonly used "mixing rules" to obtain the cross-interface force constants from bulk material force constants. Another important algorithmic development is the integration of the recursive Green's function (RGF) method with Büttiker probe scattering that enables computationally efficient simulations of inelastic phonon scattering and its contribution to the thermal interface conductance. First-principles calculations of electron-phonon coupling reveal that cross-interface energy transfer between metal electrons and atomic vibrations in the semiconductor is mediated by delocalized acoustic phonon modes that extend on both sides of the interface, and phonon modes that are localized inside the semiconductor region of the interface exhibit negligible coupling with electrons in the metal. We also provide a direct comparison between simulation predictions and experimental measurements of thermal interface conductance of epitaxial CoSi2-Si interfaces using the time-domain thermoreflectance technique. Importantly, the experimental results, performed across a wide temperature range, only agree well with

Lagrangian Transport Calculations Using UARS Data. Part I: Passive Tracers

Science.gov (United States)

Manney, G. L.; Lahoz, W. A.; Harwood, R. S.; Zurek, R. W.; Kumer, J. B.; Mergenthaler, J. L.; Roche, A. E.; O'Neill, A; Swinbank, R.; Waters, J. W.

1994-01-01

The transport of passive tracers observed by UARS has been simulated using computed trajectories of thousands of air parcels initialized on a three-dimensional stratospheric grid. These trajectories are calculated in isentropic coordinates using horizontal winds provided by the United Kingdom Meteorological Office data assimilation system and vertical (cross-isentropic) velocities computed using a fast radiation code.

CSA C873 Building Energy Estimation Methodology - A simplified monthly calculation for quick building optimization

NARCIS (Netherlands)

Legault, A.; Scott, L.; Rosemann, A.L.P.; Hopkins, M.

2014-01-01

CSA C873 Building Energy Estimation Methodology (BEEM) is a new series of (10) standards that is intended to simplify building energy calculations. The standard is based upon the German DIN Standard 18599 that has 8 years of proven track record and has been modified for the Canadian market. The BEEM

New methodology for analytical calculation of resonance integrals in an heterogeneous medium

International Nuclear Information System (INIS)

Campos, T.P.R. de; Martinez, A.S.

1986-01-01

A new methodology for analytical calculation of Resonance Integral in a typical fuel cell is presented. The expression obtained for the Resonance Integral presents the advantage of being analytical. Its constituent terms are combinations of the well known function J(xi,β) with its partial derivatives in regard to β. This is a general expression for all types of resonance. The parameters used in this method depend on the resonance type and are obtained as a function of the parameter lambda. A simple expression, depending on resonance parameters is proposed for this variable. (Author) [pt

Parallel processing of two-dimensional Sn transport calculations

International Nuclear Information System (INIS)

Uematsu, M.

1997-01-01

A parallel processing method for the two-dimensional S n transport code DOT3.5 has been developed to achieve a drastic reduction in computation time. In the proposed method, parallelization is achieved with angular domain decomposition and/or space domain decomposition. The calculational speed of parallel processing by angular domain decomposition is largely influenced by frequent communications between processing elements. To assess parallelization efficiency, sample problems with up to 32 x 32 spatial meshes were solved with a Sun workstation using the PVM message-passing library. As a result, parallel calculation using 16 processing elements, for example, was found to be nine times as fast as that with one processing element. As for parallel processing by geometry segmentation, the influence of processing element communications on computation time is small; however, discontinuity at the segment boundary degrades convergence speed. To accelerate the convergence, an alternate sweep of angular flux in conjunction with space domain decomposition and a two-step rescaling method consisting of segmentwise rescaling and ordinary pointwise rescaling have been developed. By applying the developed method, the number of iterations needed to obtain a converged flux solution was reduced by a factor of 2. As a result, parallel calculation using 16 processing elements was found to be 5.98 times as fast as the original DOT3.5 calculation

Least-cost Paths - Some Methodological Issues

Directory of Open Access Journals (Sweden)

Irmela Herzog

2014-06-01

Full Text Available This article deals with methodological issues connected with least-cost path (LCP calculations in archaeology. The number of LCP studies in archaeology has increased rapidly during the last couple of years, but not all of the approaches applied are based on an appropriate model and implementation. Many archaeologists rely on standard GIS software with default settings for calculating LCPs and are not aware of possible alternatives and the pitfalls that are described in this article. After briefly introducing the aims and applications of LCP methods in archaeology, LCP algorithms are discussed. The outcome of the LCP calculations depends not only on the algorithm but also on the cost model, which often includes several cost components. The discussion of the cost components has a focus on slope, because nearly all archaeological LCP studies take this cost component into account and because several methodological issues are connected with slope-based cost models. Other possible cost components are: the load of the walker, vegetation cover, wetlands or other soil properties, travelling and transport on water, water as barrier and as attractor, aspect, altitude, and social or cultural cost components. Eventually, advantages and disadvantages of different ways of combining cost components are presented. Based on the methodological issues I conclude that both validation checks and variations of the model are necessary to analyse the reliability of archaeological LCP results.

First principles calculations using density matrix divide-and-conquer within the SIESTA methodology

International Nuclear Information System (INIS)

Cankurtaran, B O; Gale, J D; Ford, M J

2008-01-01

The density matrix divide-and-conquer technique for the solution of Kohn-Sham density functional theory has been implemented within the framework of the SIESTA methodology. Implementation details are provided where the focus is on the scaling of the computation time and memory use, in both serial and parallel versions. We demonstrate the linear-scaling capabilities of the technique by providing ground state calculations of moderately large insulating, semiconducting and (near-) metallic systems. This linear-scaling technique has made it feasible to calculate the ground state properties of quantum systems consisting of tens of thousands of atoms with relatively modest computing resources. A comparison with the existing order-N functional minimization (Kim-Mauri-Galli) method is made between the insulating and semiconducting systems

Application of a CADIS-like variance reduction technique to electron transport

International Nuclear Information System (INIS)

Dionne, B.; Haghighat, A.

2004-01-01

This paper studies the use of approximate deterministic importance functions to calculate the lower-weight bounds of the MCNP5 weight-window variance reduction technique when applied to electron transport simulations. This approach follows the CADIS (Consistent Adjoint Driven Importance Sampling) methodology developed for neutral particles shielding calculations. The importance functions are calculated using the one-dimensional CEPXS/ONELD code package. Considering a simple 1-D problem, this paper shows that our methodology can produce speedups up to ∼82 using an approximate electron importance function distributions computed in ∼8 seconds. (author)

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock

International Nuclear Information System (INIS)

Shirakawa, Toshihiko; Hatanaka, Koichiro

2001-11-01

In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)

Calculation of Transport Coefficients in Dense Plasma Mixtures

Science.gov (United States)

Haxhimali, T.; Cabot, W. H.; Caspersen, K. J.; Greenough, J.; Miller, P. L.; Rudd, R. E.; Schwegler, E. R.

2011-10-01

We use classical molecular dynamics (MD) to estimate species diffusivity and viscosity in mixed dense plasmas. The Yukawa potential is used to describe the screened Coulomb interaction between the ions. This potential has been used widely, providing the basis for models of dense stellar materials, inertial confined plasmas, and colloidal particles in electrolytes. We calculate transport coefficients in equilibrium simulations using the Green- Kubo relation over a range of thermodynamic conditions including the viscosity and the self - diffusivity for each component of the mixture. The interdiffusivity (or mutual diffusivity) can then be related to the self-diffusivities by using a generalization of the Darken equation. We have also employed non-equilibrium MD to estimate interdiffusivity during the broadening of the interface between two regions each with a high concentration of either species. Here we present results for an asymmetric mixture between Ar and H. These can easily be extended to other plasma mixtures. A main motivation for this study is to develop accurate transport models that can be incorporated into the hydrodynamic codes to study hydrodynamic instabilities. We use classical molecular dynamics (MD) to estimate species diffusivity and viscosity in mixed dense plasmas. The Yukawa potential is used to describe the screened Coulomb interaction between the ions. This potential has been used widely, providing the basis for models of dense stellar materials, inertial confined plasmas, and colloidal particles in electrolytes. We calculate transport coefficients in equilibrium simulations using the Green- Kubo relation over a range of thermodynamic conditions including the viscosity and the self - diffusivity for each component of the mixture. The interdiffusivity (or mutual diffusivity) can then be related to the self-diffusivities by using a generalization of the Darken equation. We have also employed non-equilibrium MD to estimate interdiffusivity during

Neutron and gamma ray transport calculations in shielding system

Energy Technology Data Exchange (ETDEWEB)

Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

Neutron and photon transport calculations in fusion system. 2

Energy Technology Data Exchange (ETDEWEB)

Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1998-03-01

On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)

Hydrogen transport in a toroidal plasma using multigroup discrete-ordinates methodology

International Nuclear Information System (INIS)

Wienke, B.R.; Miller, W.F. Jr.; Seed, T.J.

1979-01-01

Neutral hydrogen transport in a fully ionized two-dimensional tokamak plasma was examined using discrete ordinates and contrasted with earlier analyses. In particular, curvature effects induced by toroidal geometries and ray effects caused by possible source localization were investigated. From an overview of the multigroup discrete-ordinates approximation, methodology in two-dimensional cylindrical geometry is detailed, mesh and plasma zoning procedures are sketched, and the piecewise polynomial solution algorithm on a triangular domain is obtained. Toroidal effects and comparisons as related to reaction rates and perticle spectra are examined for various model and source configurations

A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

Energy Technology Data Exchange (ETDEWEB)

Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Berrill, Mark A.; Allu, Srikanth; Hamilton, Steven P.; Sampath, Rahul S.; Clarno, Kevin T. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Dilts, Gary A. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

2015-04-01

This paper describes an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors is described. Details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstrating the achieved efficiency of the algorithm are presented. Furthermore, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.

Comparison of neutron transport calculations with NRC test results

International Nuclear Information System (INIS)

Koban, J.; Hofmann, W.

1981-02-01

For an exactly defined reactor arrangement (PCA = Pool Critical Assembly) neutron fluxes, neutron spectra and reaction rates for several neutron detectors were calculated by means of one and two dimensional transport codes. An international comparison proved the methods applied at KWU to be adequate. There were difficulties, however, in considering the three dimensions of the assembly which result mainly from its small dimension. This fact applies to all participants who didn't use three dimensional codes. (orig.) [de

LCA calculations on Swedish wood pellet production chains - according to the Renewable Energy Directive

Energy Technology Data Exchange (ETDEWEB)

Hagberg, Linus; Saernholm, Erik; Gode, Jenny; Ekvall, Tomas; Rydberg, Tomas

2009-09-15

The study includes calculations of typical life cycle emissions of greenhouse gases for representative Swedish pellet production chains in accordance with the calculation rules in RED (Directive 2009/28/EC). The study also intends to analyse how the directive is applicable on solid biofuels in general and on wood pellet production in particular, and to identify such aspects of the methodology in RED that are associated with obscurities, problems or lead to misleading results compared to other life cycle analysis principles. The report includes a large number of alternative calculations to show how different facts, assumptions and methodological choices affect the results. This includes the effect of what fuels are used for drying, different transport distances, assumed fuel mix for purchased electricity, the variance in efficiency between the investigated plants as well as the effect of different interpretations of the RED methodology for greenhouse gas calculations

Study of the methodology for sensitivity calculations of fast reactors integral parameters

International Nuclear Information System (INIS)

Renke, C.A.C.

1981-06-01

A study of the methodology for sensitivity calculations of integral parameters of fast reactors for the adjustment of multigroup cross sections is presented. A description of several existent methods and theories is given, with special emphasis being regarded to variational perturbation theory, integrant of the sensitivity code VARI-1D used in this work. Two calculational systems are defined and a set of procedures and criteria is structured gathering the necessary conditions for the determination of the sensitivity coefficients. These coefficients are then computed by both the direct method and the variational perturbation theory. A reasonable number of sensitivity coefficients are computed and analyzed for three fast critical assemblies, covering a range of special interest of the spectrum. These coefficients are determined for severa integral parameters, for the capture and fission cross sections of the U-238 and Pu-239, covering all the energy up to 14.5 MeV. The nuclear data used were obtained the CARNAVAL II calculational system of the Instituto de Engenharia Nuclear. An optimization for sensitivity computations in a chainned sequence of procedures is made, yielding the sensitivities in the energy macrogroups as the final stage. (Author) [pt

Relative Hazard and Risk Measure Calculation Methodology Rev 1

International Nuclear Information System (INIS)

Stenner, Robert D.; White, Michael K.; Strenge, Dennis L.; Aaberg, Rosanne L.; Andrews, William B.

2000-01-01

Documentation of the methodology used to calculate relative hazard and risk measure results for the DOE complex wide risk profiles. This methodology is used on major site risk profiles. In February 1997, the Center for Risk Excellence (CRE) was created and charged as a technical, field-based partner to the Office of Science and Risk Policy (EM-52). One of the initial charges to the CRE is to assist the sites in the development of ''site risk profiles.'' These profiles are to be relatively short summaries (periodically updated) that present a broad perspective on the major risk related challenges that face the respective site. The risk profiles are intended to serve as a high-level communication tool for interested internal and external parties to enhance the understanding of these risk-related challenges. The risk profiles for each site have been designed to qualitatively present the following information: (1) a brief overview of the site, (2) a brief discussion on the historical mission of the site, (3) a quote from the site manager indicating the site's commitment to risk management, (4) a listing of the site's top risk-related challenges, (5) a brief discussion and detailed table presenting the site's current risk picture, (6) a brief discussion and detailed table presenting the site's future risk reduction picture, and (7) graphic illustrations of the projected management of the relative hazards at the site. The graphic illustrations were included to provide the reader of the risk profiles with a high-level mental picture to associate with all the qualitative information presented in the risk profile. Inclusion of these graphic illustrations presented the CRE with the challenge of how to fold this high-level qualitative risk information into a system to produce a numeric result that would depict the relative change in hazard, associated with each major risk management action, so it could be presented graphically. This report presents the methodology developed

An investigation of fission models for high-energy radiation transport calculations

International Nuclear Information System (INIS)

Armstrong, T.W.; Cloth, P.; Filges, D.; Neef, R.D.

1983-07-01

An investigation of high-energy fission models for use in the HETC code has been made. The validation work has been directed checking the accuracy of the high-energy radiation transport computer code HETC to investigate the appropriate model for routine calculations, particularly for spallation neutron source applications. Model calculations are given in terms of neutron production, fission fragment energy release, and residual nuclei production for high-energy protons incident on thin uranium targets. The effect of the fission models on neutron production from thick uranium targets is also shown. (orig.)

THE CALCULATION OF THE ENERGY RECOVERY ELECTRIFIED URBAN TRANSPORT DURING THE INSTALLATION DRIVE FOR TRACTION SUBSTATION

Directory of Open Access Journals (Sweden)

A. A. Sulim

2014-01-01

Full Text Available At present a great attention is paid to increasing of energy efficiency at operated electrified urban transport. Perspective direction for increasing energy efficiency at that type of transport is the application of regenerative braking. For additional increasing of energy efficiency there were suggested the use of capacitive drive on tires of traction substation. One of the main task is the analysis of energy recovery application  with drive and without it.These analysis demonstrated that the calculation algorithms don’t allow in the full volume to carry out calculations of amount and cost of energy recovery without drive and with it. That is why we see the current interest to this topic. The purpose of work is to create methods of algorithms calculation for definite amount and cost of consumed, redundant and recovery energy of electrified urban transport due to definite regime of motion on wayside. There is algorithm developed, which allow to calculate amount and cost of consumed, redundant and recovery energy of electrified urban transport on wayside during the installation capacitive drive at traction substation. On the basis of developed algorithm for the definite regime of wagon motion of subway there were fulfilled the example of energy recovery amount and its cost calculation, among them with limited energy intensity drive, when there are 4 trains on wayside simultaneously.

Cobenefits of replacing car trips with alternative transportation: a review of evidence and methodological issues.

Science.gov (United States)

Xia, Ting; Zhang, Ying; Crabb, Shona; Shah, Pushan

2013-01-01

It has been reported that motor vehicle emissions contribute nearly a quarter of world energy-related greenhouse gases and cause nonnegligible air pollution primarily in urban areas. Reducing car use and increasing ecofriendly alternative transport, such as public and active transport, are efficient approaches to mitigate harmful environmental impacts caused by a large amount of vehicle use. Besides the environmental benefits of promoting alternative transport, it can also induce other health and economic benefits. At present, a number of studies have been conducted to evaluate cobenefits from greenhouse gas mitigation policies. However, relatively few have focused specifically on the transport sector. A comprehensive understanding of the multiple benefits of alternative transport could assist with policy making in the areas of transport, health, and environment. However, there is no straightforward method which could estimate cobenefits effect at one time. In this paper, the links between vehicle emissions and air quality, as well as the health and economic benefits from alternative transport use, are considered, and methodological issues relating to the modelling of these cobenefits are discussed.

Cobenefits of Replacing Car Trips with Alternative Transportation: A Review of Evidence and Methodological Issues

Directory of Open Access Journals (Sweden)

Ting Xia

2013-01-01

Full Text Available It has been reported that motor vehicle emissions contribute nearly a quarter of world energy-related greenhouse gases and cause nonnegligible air pollution primarily in urban areas. Reducing car use and increasing ecofriendly alternative transport, such as public and active transport, are efficient approaches to mitigate harmful environmental impacts caused by a large amount of vehicle use. Besides the environmental benefits of promoting alternative transport, it can also induce other health and economic benefits. At present, a number of studies have been conducted to evaluate cobenefits from greenhouse gas mitigation policies. However, relatively few have focused specifically on the transport sector. A comprehensive understanding of the multiple benefits of alternative transport could assist with policy making in the areas of transport, health, and environment. However, there is no straightforward method which could estimate cobenefits effect at one time. In this paper, the links between vehicle emissions and air quality, as well as the health and economic benefits from alternative transport use, are considered, and methodological issues relating to the modelling of these cobenefits are discussed.

Improvement in decay ratio calculation in LAPUR5 methodology for BWR instability

International Nuclear Information System (INIS)

Li Hsuannien; Yang Tzungshiue; Shih Chunkuan; Wang Jongrong; Lin Haotzu

2009-01-01

LAPUR5, based on frequency domain approach, is a computer code that analyzes the core stability and calculates decay ratios (DRs) of boiling water nuclear reactors. In current methodology, one set of parameters (three friction multipliers and one density reactivity coefficient multiplier) is chosen for LAPUR5 input files, LAPURX and LAPURW. The calculation stops and DR for this particular set of parameters is obtained when the convergence criteria (pressure, mass flow rate) are first met. However, there are other sets of parameters which could also meet the same convergence criteria without being identified. In order to cover these ranges of parameters, we developed an improved procedure to calculate DR in LAPUR5. First, we define the ranges and increments of those dominant input parameters in the input files for DR loop search. After LAPUR5 program execution, we can obtain all DRs for every set of parameters which satisfy the converge criteria in one single operation. The part for loop search procedure covers those steps in preparing LAPURX and LAPURW input files. As a demonstration, we looked into the reload design of Kuosheng Unit 2 Cycle 22. We found that the global DR has a maximum at exposure of 9070 MWd/t and the regional DR has a maximum at exposure of 5770 MWd/t. It should be noted that the regional DR turns out to be larger than the global ones for exposures less than 5770 MWd/t. Furthermore, we see that either global or regional DR by the loop search method is greater than the corresponding values from our previous approach. It is concluded that the loop search method can reduce human error and save human labor as compared with the previous version of LAPUR5 methodology. Now the maximum DR can be effectively obtained for a given plant operating conditions and a more precise stability boundary, with less uncertainty, can be plotted on plant power/flow map. (author)

Status of shielding analysis methods for transport packages

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.; Brady, M.C.

1991-01-01

Shielding analysis methods for transport packages are becoming more important to the cask designer because optimized cask designs with higher payloads can yield doses near the limits set by regulatory authorities. Uncertainty arising from generation of radiation sources, selection of cross-section data, and the radiation transport methodology must be considered. Recent comparison studies using popular US codes illustrate calculational discrepancies arising from each of these areas

Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

Science.gov (United States)

Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko

2017-09-01

In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.

Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

International Nuclear Information System (INIS)

Downar, T.

2009-01-01

The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.

Two dimensional neutron transport calculation system for plate-reactors: experimental design and qualification with SILOE

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

Methodology for worker neutron exposure evaluation in the PDCF facility design

International Nuclear Information System (INIS)

Scherpelz, R. I.; Traub, R. J.; Pryor, K. H.

2004-01-01

A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y -1 for the whole body and 100 mSv y -1 for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were

MODEL OF FEES CALCULATION FOR ACCESS TO TRACK INFRASTRUCTURE FACILITIES

Directory of Open Access Journals (Sweden)

M. I. Mishchenko

2014-12-01

Full Text Available Purpose. The purpose of the article is to develop a one- and two-element model of the fees calculation for the use of track infrastructure of Ukrainian railway transport. Methodology. On the basis of this one can consider that when planning the planned preventive track repair works and the amount of depreciation charges the guiding criterion is not the amount of progress it is the operating life of the track infrastructure facilities. The cost of PPTRW is determined on the basis of the following: the classification track repairs; typical technological processes for track repairs; technology based time standards for PPTRW; costs for the work of people, performing the PPTRW, their hourly wage rates according to the Order 98-Ts; the operating cost of machinery; regulated list; norms of expenditures and costs of materials and products (they have the largest share of the costs for repairs; railway rates; average distances for transportation of materials used during repair; standards of general production expenses and the administrative costs. Findings. The models offered in article allow executing the objective account of expenses in travelling facilities for the purpose of calculation of the proved size of indemnification and necessary size of profit, the sufficient enterprises for effective activity of a travelling infrastructure. Originality. The methodological bases of determination the fees (payments for the use of track infrastructure on one- and two-element base taking into account the experience of railways in the EC countries and the current transport legislation were grounded. Practical value. The article proposes the one- and two-element models of calculating the fees (payments for the TIF use, accounting the applicable requirements of European transport legislation, which provides the expense compensation and income formation, sufficient for economic incentives of the efficient operation of the TIE of Ukrainian railway transport.

Application of the API/NPRA SVA methodology to transportation security issues.

Science.gov (United States)

Moore, David A

2006-03-17

Security vulnerability analysis (SVA) is becoming more prevalent as the issue of chemical process security is of greater concern. The American Petroleum Institute (API) and the National Petrochemical and Refiner's Association (NPRA) have developed a guideline for conducting SVAs of petroleum and petrochemical facilities in May 2003. In 2004, the same organizations enhanced the guidelines by adding the ability to evaluate transportation security risks (pipeline, truck, and rail). The importance of including transportation and value chain security in addition to fixed facility security in a SVA is that these issues may be critically important to understanding the total risk of the operation. Most of the SVAs done using the API/NPRA SVA and other SVA methods were centered on the fixed facility and the operations within the plant fence. Transportation interfaces alone are normally studied as a part of the facility SVA, and the entire transportation route impacts and value chain disruption are not commonly considered. Particularly from a national, regional, or local infrastructure analysis standpoint, understanding the interdependencies is critical to the risk assessment. Transportation risks may include weaponization of the asset by direct attack en route, sabotage, or a Trojan Horse style attack into a facility. The risks differ in the level of access control and the degree of public exposures, as well as the dynamic nature of the assets. The public exposures along the transportation route need to be carefully considered. Risks may be mitigated by one of many strategies including internment, staging, prioritization, conscription, or prohibition, as well as by administrative security measures and technology for monitoring and isolating the assets. This paper illustrates how these risks can be analyzed by the API/NPRA SVA methodology. Examples are given of a pipeline operation, and other examples are found in the guidelines.

Improvement of the efficiency of two-dimensional multigroup transport calculations assuming isotropic reflection with multilevel spatial discretisation

International Nuclear Information System (INIS)

Stankovski, Z.; Zmijarevic, I.

1987-06-01

This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy

Development of probabilistic assessment methodology for geologic disposal of radioactive wastes

International Nuclear Information System (INIS)

Kimura, H.; Takahashi, T.

1998-01-01

The probabilistic assessment methodology is essential to evaluate uncertainties of long-term radiological consequences associated with geologic disposal of radioactive wastes. We have developed a probabilistic assessment methodology to estimate the influences of parameter uncertainties/variabilities. An exposure scenario considered here is based on a groundwater migration scenario. A computer code system GSRW-PSA thus developed is based on a non site-specific model, and consists of a set of sub modules for sampling of model parameters, calculating the release of radionuclides from engineered barriers, calculating the transport of radionuclides through the geosphere, calculating radiation exposures of the public, and calculating the statistical values relating the uncertainties and sensitivities. The results of uncertainty analyses for α-nuclides quantitatively indicate that natural uranium ( 238 U) concentration is suitable for an alternative safety indicator of long-lived radioactive waste disposal, because the estimated range of individual dose equivalent due to 238 U decay chain is narrower that that due to other decay chain ( 237 Np decay chain). It is internationally necessary to have detailed discussion on the PDF of model parameters and the PSA methodology to evaluated the uncertainties due to conceptual models and scenarios. (author)

Repair for scattering expansion truncation errors in transport calculations

International Nuclear Information System (INIS)

Emmett, M.B.; Childs, R.L.; Rhoades, W.A.

1980-01-01

Legendre expansion of angular scattering distributions is usually limited to P 3 in practical transport calculations. This truncation often results in non-trivial errors, especially alternating negative and positive lateral scattering peaks. The effect is especially prominent in forward-peaked situations such as the within-group component of the Compton Scattering of gammas. Increasing the expansion to P 7 often makes the peaks larger and narrower. Ward demonstrated an accurate repair, but his method requires special cross section sets and codes. The DOT IV code provides fully-compatible, but heuristic, repair of the erroneous scattering. An analytical Klein-Nishina estimator, newly available in the MORSE code, allows a test of this method. In the MORSE calculation, particle scattering histories are calculated in the usual way, with scoring by an estimator routine at each collision site. Results for both the conventional P 3 estimator and the analytical estimator were obtained. In the DOT calculation, the source moments are expanded into the directional representation at each iteration. Optionally a sorting procedure removes all negatives, and removes enough small positive values to restore particle conservation. The effect of this is to replace the alternating positive and negative values with positive values of plausible magnitude. The accuracy of those values is examined herein

RAMA Methodology for the Calculation of Neutron Fluence; Metodologia RAMA para el Calculo de la Fluencia Neutronica

Energy Technology Data Exchange (ETDEWEB)

Villescas, G.; Corchon, F.

2013-07-01

he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.

A proposed methodology for performing risk analysis of state radiation control programs

International Nuclear Information System (INIS)

Dornsife, W.P.

1996-01-01

This paper is comprised of viewgraphs from a conference presentation. Topics discussed include barriers to effective risk assessment and management, and real versus perceived risk for various radiation programs in the state of Pennsylvania. Calculation results for Pennsylvania are provided for low-level radioactive waste transportation risks, indoor radon risk, and cancer morbidity risk from x-rays. A methodology for prioritizing radiation regulatory programs based on risk is presented with calculations for various Pennsylvania programs

Theoretical prediction of the electronic transport properties of the Al-Cu alloys based on the first-principle calculation and Boltzmann transport equation

Science.gov (United States)

Choi, Garam; Lee, Won Bo

Metal alloys, especially Al-based, are commonly-used materials for various industrial applications. In this paper, the Al-Cu alloys with varying the Al-Cu ratio were investigated based on the first-principle calculation using density functional theory. And the electronic transport properties of the Al-Cu alloys were carried out using Boltzmann transport theory. From the results, the transport properties decrease with Cu-containing ratio at the temperature from moderate to high, but with non-linearity. It is inferred by various scattering effects from the calculation results with relaxation time approximation. For the Al-Cu alloy system, where it is hard to find the reliable experimental data for various alloys, it supports understanding and expectation for the thermal electrical properties from the theoretical prediction. Theoretical and computational soft matters laboratory.

Method for calculating anisotropic neutron transport using scattering kernel without polynomial expansion

International Nuclear Information System (INIS)

Takahashi, Akito; Yamamoto, Junji; Ebisuya, Mituo; Sumita, Kenji

1979-01-01

A new method for calculating the anisotropic neutron transport is proposed for the angular spectral analysis of D-T fusion reactor neutronics. The method is based on the transport equation with new type of anisotropic scattering kernels formulated by a single function I sub(i) (μ', μ) instead of polynomial expansion, for instance, Legendre polynomials. In the calculation of angular flux spectra by using scattering kernels with the Legendre polynomial expansion, we often observe the oscillation with negative flux. But in principle this oscillation disappears by this new method. In this work, we discussed anisotropic scattering kernels of the elastic scattering and the inelastic scatterings which excite discrete energy levels. The other scatterings were included in isotropic scattering kernels. An approximation method, with use of the first collision source written by the I sub(i) (μ', μ) function, was introduced to attenuate the ''oscillations'' when we are obliged to use the scattering kernels with the Legendre polynomial expansion. Calculated results with this approximation showed remarkable improvement for the analysis of the angular flux spectra in a slab system of lithium metal with the D-T neutron source. (author)

An efficient methodology of two groups spatial calculation for neutronic state and sensisivity coefficients in fast reactors

International Nuclear Information System (INIS)

Jachic, J.

1985-01-01

It is presented the ONEDM neutronic simulator for RZ spatial calculation, two energy groups, aiming at researching and optimization of a low power fast reactor design. The simulator's methodology is based in RZ calculation from radial and axial calculation iteractively coupled and in macroscopic cross sections corrected by power density and asymmetry of the spectrum in the feedback process with phase library for reference neutronic state. The transversal area which are determined by energy groups and material region in the iteration are introduced in the spatial calculation. The simulator efficiency is tested and compared with the CITATION and 2DB codes. The cross sections are generated by 1DX code. (M.C.K.) [pt

Hybrid PN-SN Calculations with SAAF for the Multiscale Transport Capability in Rattlesnake

Energy Technology Data Exchange (ETDEWEB)

Wang, Yaqi; Schunert, Sebastian; DeHart, Mark; Martineau, Richard

2016-05-01

Two interface conditions, the Lagrange multiplier method and the upwinding method, for hybrid \\pn-\\sn calculations is proposed for the self-adjoint angular flux (SAAF) formulation of the transport equation using the continuous finite element method (FEM) for spatial discretization. These interface conditions are implemented in Rattlesnake, the radiation transport application built on MOOSE, for the on-going multiscale transport simulation effort at INL. For smoothing the solution at the interface for the Lagrange multiplier method, a method based on \\sn Lagrange interpolation on the sphere is proposed. Numerical results indicate that the interface conditions give the expected convergence.

Hyperfine 3D neutronic calculations in CANDU supercells

International Nuclear Information System (INIS)

Balaceanu, V.; Aioanei, L.; Pavelescu, M.

2010-01-01

For an accurate evaluation of the fuel performances, it is very important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle. According this issue, in our Institute, a multigroup calculation methodology named WIMS-PIJXYZ was especially developed for estimating the local neutronic parameters in CANDU cell/supercells. The objective of this paper is to present this calculation methodology and to use it in performing some hyperfine neutronic calculations in CANDU type supercells. More exactly, after a short description for the WIMS-PIJXYZ methodology, the end effect for some CANDU fuel bundles is estimated. The WIMS-PIJXYZ methodology is based on WIMS and PIJXYZ transport codes. WIMS is a standard lattice-cell code and it is used for generating the multigroup macroscopic cross sections for the materials in the fuel cells. For obtaining the flux and power distributions in CANDU fuel bundles the PIJXYZ code is used. This code is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The end effect consists in the increasing of the thermal neutron flux in the end region and the increasing of power in the end of the fuel rod. The region separating the CANDU fuel in two adjoining bundles in a channel is called the 'end region' and the end of the last pellet in the fuel stack adjacent to the end region is called the 'fuel end'. The end effect appears because the end region of the bundle is made up of coolant and Zircaloy-4, a very low neutron absorption material. To estimate the end effect, the flux peaking factors and the power peaking factors are calculated. It was taken in consideration CANDU Standard (Natural Uranium, with 37 elements) fuel bundles. In the end of the paper, the results obtained by WIMS-PIJXYZ methodology with the similar LEGENTR results are compared. The comparative analysis shows a good agreement. (authors)

Search for a transport method for the calculation of the PWR control and safety clusters

International Nuclear Information System (INIS)

Bruna, G.B.; Van Frank, C.; Vergain, M.L.; Chauvin, J.P.; Palmiotti, G.; Nobile, M.

1990-01-01

The project studies of power reactors rely mainly on diffusion calculations, but transport ones are often needed for assessing fine effects, intimately linked to geometry and spectrum heterogeneities. Accurate transport computations are necessary, in particular, for shielded cross section generation, and when homogenization and dishomogenization processes are involved. The transport codes, generally, offer the user a variety of computational options, related to different approximation levels. In every case, it is obviously desirable to be able to choose the reliable degree of approximation to be accepted in any particular computational circumstance of the project. The search for such adapted procedures is to be made on the basis of critical experiments. In our studies, this task was made possible by the availability of suitable results of the CAMELEON critical experiment, carried on in the EOLE facility at CEA's Center of Cadarache. In this paper, we summarize some of the work in progress at FRAMATOME on the definition of an assembly based transport calculation scheme to be used for PWR control and safety cluster computations. Two main items, devoted to the search of the optimum computational procedures, are presented here: - a parametrical study on computational options, made in an infinite medium assembly geometry, - a series of comparisons between calculated and experimental values of pin power distribution

An accurate solver for forward and inverse transport

International Nuclear Information System (INIS)

Monard, Francois; Bal, Guillaume

2010-01-01

This paper presents a robust and accurate way to solve steady-state linear transport (radiative transfer) equations numerically. Our main objective is to address the inverse transport problem, in which the optical parameters of a domain of interest are reconstructed from measurements performed at the domain's boundary. This inverse problem has important applications in medical and geophysical imaging, and more generally in any field involving high frequency waves or particles propagating in scattering environments. Stable solutions of the inverse transport problem require that the singularities of the measurement operator, which maps the optical parameters to the available measurements, be captured with sufficient accuracy. This in turn requires that the free propagation of particles be calculated with care, which is a difficult problem on a Cartesian grid. A standard discrete ordinates method is used for the direction of propagation of the particles. Our methodology to address spatial discretization is based on rotating the computational domain so that each direction of propagation is always aligned with one of the grid axes. Rotations are performed in the Fourier domain to achieve spectral accuracy. The numerical dispersion of the propagating particles is therefore minimal. As a result, the ballistic and single scattering components of the transport solution are calculated robustly and accurately. Physical blurring effects, such as small angular diffusion, are also incorporated into the numerical tool. Forward and inverse calculations performed in a two-dimensional setting exemplify the capabilities of the method. Although the methodology might not be the fastest way to solve transport equations, its physical accuracy provides us with a numerical tool to assess what can and cannot be reconstructed in inverse transport theory.

Simplified methodology for control cell constant calculations of the reactor cores for the space kinetics

International Nuclear Information System (INIS)

Santos, Rubens Souza dos; Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques

2002-01-01

In this work is presented a methodology which focuses the distribution of neutron absorber rods in nuclear reactor power plants, for utilizing in space kinetic calculations, principally in the cluster ejection transients of control rods. A numerical model for macroscopic constant calculations based on the knowledge of the neutron flux without the control rods is proposed, as alternative to the analytical models, based on the hypothesis of the null current on the cell super boundaries. The proposed model in this work has itself showed adequate to deal with problems with strong space dependence, once that the model showed consistence in the global average built in the analytical model. (author)

A systematic framework for effective uncertainty assessment of severe accident calculations; Hybrid qualitative and quantitative methodology

International Nuclear Information System (INIS)

Hoseyni, Seyed Mohsen; Pourgol-Mohammad, Mohammad; Tehranifard, Ali Abbaspour; Yousefpour, Faramarz

2014-01-01

This paper describes a systematic framework for characterizing important phenomena and quantifying the degree of contribution of each parameter to the output in severe accident uncertainty assessment. The proposed methodology comprises qualitative as well as quantitative phases. The qualitative part so called Modified PIRT, being a robust process of PIRT for more precise quantification of uncertainties, is a two step process for identifying and ranking based on uncertainty importance in severe accident phenomena. In this process identified severe accident phenomena are ranked according to their effect on the figure of merit and their level of knowledge. Analytical Hierarchical Process (AHP) serves here as a systematic approach for severe accident phenomena ranking. Formal uncertainty importance technique is used to estimate the degree of credibility of the severe accident model(s) used to represent the important phenomena. The methodology uses subjective justification by evaluating available information and data from experiments, and code predictions for this step. The quantitative part utilizes uncertainty importance measures for the quantification of the effect of each input parameter to the output uncertainty. A response surface fitting approach is proposed for estimating associated uncertainties with less calculation cost. The quantitative results are used to plan in reducing epistemic uncertainty in the output variable(s). The application of the proposed methodology is demonstrated for the ACRR MP-2 severe accident test facility. - Highlights: • A two stage framework for severe accident uncertainty analysis is proposed. • Modified PIRT qualitatively identifies and ranks uncertainty sources more precisely. • Uncertainty importance measure quantitatively calculates effect of each uncertainty source. • Methodology is applied successfully on ACRR MP-2 severe accident test facility

The effect of gamma-ray transport on afterheat calculations for accident analysis

International Nuclear Information System (INIS)

Reyes, S.; Latkowski, J.F.; Sanz, J.

2000-01-01

Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed

Error reduction techniques for Monte Carlo neutron transport calculations

International Nuclear Information System (INIS)

Ju, J.H.W.

1981-01-01

Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

Calculation of the coherent transport properties of a symmetric spin nanocontact

International Nuclear Information System (INIS)

Bourahla, B.; Khater, A.; Tigrine, R.

2009-01-01

A theoretical study is presented for the coherent transport properties of a magnetic nanocontact. In particular, we study a symmetric nanocontact between two identical waveguides composed of semi-infinite spin ordered ferromagnetic chains. The coherent transmission and reflection scattering cross sections via the nanocontact, for spin waves incident from the bulk waveguide, are calculated with the use of the matching method. The inter-atomic magnetic exchange on the nanocontact is allowed to vary to investigate the consequences of magnetic softening and hardening for the calculated spectra. Transmission spectra underline the filtering properties of the nanocontact. The localized spin density of states in the nanocontact domain is also calculated, and analyzed. The results yield an understanding of the relationship between coherent conductance and the structural configuration of the nanocontact.

Study on the Development of Methodology for Cost Calculations and Financial Planning of Decommissioning Operations

International Nuclear Information System (INIS)

2001-12-01

The following study deals with the development of methodology for cost calculations and financial planning of decommissioning operations. It has been carried out by EDF / FRAMATOME / VUJE / SCK-CEN in the frame of the contract B7-032/2000/291058/MAR/C2 awarded by the European Commission. This study consists of 4 parts. The first task objective is to develop a reliable and transparent methodology for cost assessment and financial planning sufficient precise but without long and in depth investigations and studies. This methodology mainly contains: Calculation methods and algorithms for the elaboration of costs items making up the whole decommissioning cost. Estimated or standard values for the parameters and for the cost factors to be used in the above-mentioned algorithms Financial mechanism to be applied as to establish a financial planning. The second part task is the provision of standard values for the different parameters and costs factors described in the above-mentioned algorithms. This provision of data is based on the own various experience acquired by the members of the working team and on existing international references (databases, publications and reports). As decommissioning operations are spreading over several dozens of years, the scope of this task the description of the financial mechanisms to be applied to the different cost items as to establish a complete financial cost. It takes into account the financial schedule issued in task 1. The scope of this task consists in bringing together in a guideline all the information collected before: algorithms, data and financial mechanisms. (A.L.B.)

A method for local transport analysis in tokamaks with error calculation

International Nuclear Information System (INIS)

Hogeweij, G.M.D.; Hordosy, G.; Lopes Cardozo, N.J.

1989-01-01

Global transport studies have revealed that heat transport in a tokamak is anomalous, but cannot provide information about the nature of the anomaly. Therefore, local transport analysis is essential for the study of anomalous transport. However, the determination of local transport coefficients is not a trivial affair. Generally speaking one can either directly measure the heat diffusivity, χ, by means of heat pulse propagation analysis, or deduce the profile of χ from measurements of the profiles of the temperature, T, and the power deposition. Here we are concerned only with the latter method, the local power balance analysis. For the sake of clarity heat diffusion only is considered: ρ=-gradT/q (1) where ρ=κ -1 =(nχ) -1 is the heat resistivity and q is the heat flux per unit area. It is assumed that the profiles T(r) and q(r) are given with some experimental error. In practice T(r) is measured directly, e.g. from ECE spectroscopy, while q(r) is deduced from the power deposition and loss profiles. The latter cannot be measured directly and is partly determined on the basis of models. This complication will not be considered here. Since in eq. (1) the gradient of T appears, noise on T can severely affect the solution ρ. This means that in general some form of smoothing must be applied. A criterion is needed to select the optimal smoothing. Too much smoothing will wipe out the details, whereas with too little smoothing the noise will distort the reconstructed profile of ρ. Here a new method to solve eq. (1) is presented which expresses ρ(r) as a cosine-series. The coefficients of this series are given as linear combinations of the Fourier coefficients of the measured T- and q-profiles. This formulation allows 1) the stable and accurate calculation of the ρ-profile, and 2) the analytical calculation of the error in this profile. (author) 5 refs., 3 figs

Methodology for calculation of carbon emission and energy generation efficiency by fossil coal thermal power plants

International Nuclear Information System (INIS)

Licks, Leticia A.; Pires, Marcal

2008-01-01

This work intends to evaluate the emissions of carbon dioxide (CO 2 ) emitted by the burning of fossil coal in Brazil. So, a detailed methodology is proposed for calculation of CO 2 emissions from the carbon emission coefficients specific for the Brazilian carbons. Also, the using of secondary fuels (fuel oil and diesel oil) were considered and the power generation for the calculation of emissions and efficiencies of each power plant as well. The obtained results indicate carbon emissions for the year 2002 approximately of the order of 1,794 Gg, with 20% less than the obtained by the official methodology (MCT). Such differences are related to the non consideration of the humidity containment of the coals as well as the using of generic coefficients not adapted to the Brazilian coals. The obtained results indicate the necessity to review the emission inventories and the modernization of the burning systems aiming the increase the efficiency and reduction of the CO 2 and other pollutants, as an alternative for maintaining the sustainable form of using the fossil coal in the country

Structural instability of atmospheric flows under perturbations of the mass balance and effect in transport calculations

International Nuclear Information System (INIS)

Núñez, M A; Mendoza, R

2015-01-01

Several methods to estimate the velocity field of atmospheric flows, have been proposed to the date for applications such as emergency response systems, transport calculations and for budget studies of all kinds. These applications require a wind field that satisfies the conservation of mass but, in general, estimated wind fields do not satisfy exactly the continuity equation. An approach to reduce the effect of using a divergent wind field as input in the transport-diffusion equations, was proposed in the literature. In this work, a linear local analysis of a wind field, is used to show analytically that the perturbation of a large-scale nondivergent flow can yield a divergent flow with a substantially different structure. The effects of these structural changes in transport calculations are illustrated by means of analytic solutions of the transport equation

New nonlinear methods for linear transport calculations

International Nuclear Information System (INIS)

Adams, M.L.

1993-01-01

We present a new family of methods for the numerical solution of the linear transport equation. With these methods an iteration consists of an 'S N sweep' followed by an 'S 2 -like' calculation. We show, by analysis as well as numerical results, that iterative convergence is always rapid. We show that this rapid convergence does not depend on a consistent discretization of the S 2 -like equations - they can be discretized independently from the S N equations. We show further that independent discretizations can offer significant advantages over consistent ones. In particular, we find that in a wide range of problems, an accurate discretization of the S 2 -like equation can be combined with a crude discretization of the S N equations to produce an accurate S N answer. We demonstrate this by analysis as well as numerical results. (orig.)

Using the WIMS-DIREN bigroup and multigroup methodology for Cernavoda Unit 1 and Unit 2 adjuster rods comparative reactivity calculations at Phase B commissioning

International Nuclear Information System (INIS)

Prodea, Iosif; Patrulescu, Ilie; Rizoiu, Andrei; Danila, Nicolae; Prisecaru, Ilie

2007-01-01

One of the most important CANDU reactor regulation system is the Adjuster Rods System (ADJ). The individual and bank calibration and performance evaluation of this system is carried out during the Phase B commissioning. The ADJ rods are grouped into seven banks based on full power reactivity control requirements. The Cernavoda Unit 2 adjuster rods characteristics were designed more than twenty years ago at INR Pitesti in the end of a fruitful collaboration between INR Pitesti (as designer) and Bristol Aerospace Limited (as manufacturer). In 1996, during the Phase B commissioning tests only AECL diffusion and Westcott approximation methodology was used. An alternative integral transport and high-modes diffusion approximation methodology was developed in INR Pitesti during the last years. As a result, the first collision probability code PIJXYZ was created and developed to carry out the supercell calculations as well as the code DIREN for 3D diffusion-based core simulations. The aim of this work was to evaluate comparatively the two adjuster rods systems (from Unit 1 and 2) in commissioning conditions. The concrete results will consist of individual, bank and total adjuster rods reactivity estimations with an emphasis on the differences and similarities between them. (authors)

Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

International Nuclear Information System (INIS)

Garcia-Herranz, Nuria; Cabellos, Oscar; Sanz, Javier; Juan, Jesus; Kuijper, Jim C.

2008-01-01

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files

Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

Energy Technology Data Exchange (ETDEWEB)

Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

2008-04-15

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

A dynamic management of a public transportation fleet

Directory of Open Access Journals (Sweden)

Ireneusz Celiński

2013-09-01

Full Text Available Background: The present paper deals with the problems of a public transportation fleet management in public transportation operators. A management concept is proposed based on a real-time acquisition of parameters of public transportation passenger exchange. Methods: The relevant research utilised video materials documenting the processes of passenger exchange in public transportation. The proposed methodology is based on a dynamic real-time measurement of passenger streams. A characteristic feature of the measurement methodology applied is that the data is collected outside the vehicles, with a CCTV camera used per access point. Demand for the public transportation service are calculated using the image processing. Results: The derived demand characteristics allow not only an estimation of the magnitude of traffic streams in public transportation but also their qualitative description. Such an approach permits a flexible design of the transportation offer to adapt to the demand. This allows matching the timetables to the density functions describing the demand for public transportation within the space of transportation networks. In addition, based on the results of this type of research, a public transportation operator may despatch the vehicle base in a flexible way. For each run of a bus or tram fleet, basing on the registered passenger traffic streams, it is possible to rationally despatch the vehicles with suitable capacity. Conclusions: A system of this type is capable of determining the quality of work of the public transportation. With the ITS systems being introduced still more widely, the proposed methodology allows the design and implementation of dynamic timetables.

Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

International Nuclear Information System (INIS)

Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

2015-01-01

The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

Development of a design methodology for hydraulic pipelines carrying rectangular capsules

International Nuclear Information System (INIS)

Asim, Taimoor; Mishra, Rakesh; Abushaala, Sufyan; Jain, Anuj

2016-01-01

The scarcity of fossil fuels is affecting the efficiency of established modes of cargo transport within the transportation industry. Efforts have been made to develop innovative modes of transport that can be adopted for economic and environmental friendly operating systems. Solid material, for instance, can be packed in rectangular containers (commonly known as capsules), which can then be transported in different concentrations very effectively using the fluid energy in pipelines. For economical and efficient design of such systems, both the local flow characteristics and the global performance parameters need to be carefully investigated. Published literature is severely limited in establishing the effects of local flow features on system characteristics of Hydraulic Capsule Pipelines (HCPs). The present study focuses on using a well validated Computational Fluid Dynamics (CFD) tool to numerically simulate the solid-liquid mixture flow in both on-shore and off-shore HCPs applications including bends. Discrete Phase Modelling (DPM) has been employed to calculate the velocity of the rectangular capsules. Numerical predictions have been used to develop novel semi-empirical prediction models for pressure drop in HCPs, which have then been embedded into a robust and user-friendly pipeline optimisation methodology based on Least-Cost Principle. - Highlights: • Local flow characteristics in a pipeline transporting rectangular capsules. • Development of prediction models for the pressure drop contribution of capsules. • Methodology developed for sizing of Hydraulic Capsule Pipelines. • Implementation of the developed methodology to obtain optimal pipeline diameter.

Calculation of the isotope concentrations, source terms and radiation shielding of the SAFARI-1 irradiation products

International Nuclear Information System (INIS)

Stoker, C.C.; Ball, G.

2000-01-01

The ever increasing expansion of the irradiation product portfolio of the SAFARI-1 reactor leads to the need to routinely calculate the radio-isotope concentrations and source terms for the materials irradiated in the reactor accurately. In addition to this, the required shielding for the transportation and processing of these irradiation products needs to be determined. In this paper the calculational methodology applied is described with special attention given to the spectrum dependence of the one-group cross sections of selected SAFARI-1 irradiation materials and the consequent effect on the determination of the isotope concentrations and source terms. Comparisons of the calculated isotopic concentrations and dose rates with experimental analysis and measurements provide confidence in the calculational methodologies and data used. (author)

Transport calculations for a 14.8 MeV neutron beam in a water phantom

International Nuclear Information System (INIS)

Goetsch, S.J.

1981-01-01

A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented

Calculation of health risks from spent-nuclear-fuel transportation accidents

International Nuclear Information System (INIS)

Chen, S.Y.; Yuan, Y.C.

1987-01-01

Models developed to analyze potential radiological health risks from various accident scenarios during transportation of spent nuclear fuels are described. The models are designed both for detailed route-specific risk analyses and for use in conducting overall risk analyses for route selection and related decision-making activities. The radiological risks calculated include individual dose commitments, collective dose commitments, and long-term (100-year) environmental dose commitments to a population following release of radioactivity. To facilitate route-specific analysis, a state-level database was developed and incorporated into the model. Route-specific analysis is demonstrated by the calculation of radiological risks resulting from various accident scenarios, as postulated by the recent US Nuclear Regulatory Commission Modal Study, for four representative states selected from various regions of the United States. 10 refs., 3 figs., 3 tabs

Activity-Based Costing Application in an Urban Mass Transport Company

Directory of Open Access Journals (Sweden)

Popesko Boris

2011-12-01

Full Text Available The purpose of this paper is to provide a basic overview of the application of Activity-Based Costing in an urban mass transport company which operates land public transport via buses and trolleys within the city. The case study was conducted using the Activity-Based Methodology in order to calculate the true cost of individual operations and to measure the profitability of particular transport lines. The case study analysis showed the possible effects of the application of the Activity-Based Costing for an urban mass transport company as well as the limitations of using the ABC methodology in the service industry. With regards to the application of the ABC methodology, the primary limitation of the accuracy of the conclusions is the quality of the non-financial information which had to be gathered throughout the implementation process. A basic limitation of the accurate data acquisition is the nature of the fare system of the transport company which does not allow the identification of the route that is taken by an individual passenger. The study illustrates the technique of ABC in urban mass transport and provides a real company example of information outputs of the ABC system. The users indicated that, the ABC model is very useful for profitability reporting and profit management. Also, the paper shows specific application of the Activity-Based Methodology in conditions of urban mass transport companies with regional specifics.

Comparison of the ESTRO formalism for monitor unit calculation with a Clarkson based algorithm of a treatment planning system and a traditional ''full-scatter'' methodology

International Nuclear Information System (INIS)

Pirotta, M.; Aquilina, D.; Bhikha, T.; Georg, D.

2005-01-01

The ESTRO formalism for monitor unit (MU) calculations was evaluated and implemented to replace a previous methodology based on dosimetric data measured in a full-scatter phantom. This traditional method relies on data normalised at the depth of dose maximum (z m ), as well as on the utilisation of the BJR 25 table for the conversion of rectangular fields into equivalent square fields. The treatment planning system (TPS) was subsequently updated to reflect the new beam data normalised at a depth z R of 10 cm. Comparisons were then carried out between the ESTRO formalism, the Clarkson-based dose calculation algorithm on the TPS (with beam data normalised at z m and z R ), and the traditional ''full-scatter'' methodology. All methodologies, except for the ''full-scatter'' methodology, separated head-scatter from phantom-scatter effects and none of the methodologies; except for the ESTRO formalism, utilised wedge depth dose information for calculations. The accuracy of MU calculations was verified against measurements in a homogeneous phantom for square and rectangular open and wedged fields, as well as blocked open and wedged fields, at 5, 10, and 20 cm depths, under fixed SSD and isocentric geometries for 6 and 10 MV. Overall, the ESTRO Formalism showed the most accurate performance, with the root mean square (RMS) error with respect to measurements remaining below 1% even for the most complex beam set-ups investigated. The RMS error for the TPS deteriorated with the introduction of a wedge, with a worse RMS error for the beam data normalised at z m (4% at 6 MV and 1.6% at 10 MV) than at z R (1.9% at 6 MV and 1.1% at 10 MV). The further addition of blocking had only a marginal impact on the accuracy of this methodology. The ''full-scatter'' methodology showed a loss in accuracy for calculations involving either wedges or blocking, and performed worst for blocked wedged fields (RMS errors of 7.1% at 6 MV and 5% at 10 MV). The origins of these discrepancies were

Transport and hydrodynamic calculations of direct photons at FAIR

International Nuclear Information System (INIS)

Baeuchle, Bjorn; Bleicher, Marcus

2011-01-01

The microscopic transport model UrQMD and a micro + macro hybrid model are used to calculate direct photon spectra from U+U-collisions at E lab =35 A GeV as will be measured by the CBM Collaboration at FAIR. In the hybrid model, the intermediate high-density part of the nuclear interaction is described with ideal 3+1-dimensional hydrodynamics. Different equations of state of the matter created in the heavy-ion collisions are investigated and the resulting spectra of direct photons are predicted. The emission patterns of direct photons in space and time are discussed.

Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

International Nuclear Information System (INIS)

Bernnat, W.; Keinert, J.; Mattes, M.

2004-01-01

For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H 2 O, liquid He, liquid D 2 O, liquid and solid H 2 and D 2 , solid CH 4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S N -transport codes and the Monte Carlo Code MCNP. (orig.)

Towards quantitative accuracy in first-principles transport calculations: The GW method applied to alkane/gold junctions

DEFF Research Database (Denmark)

Strange, Mikkel; Thygesen, Kristian Sommer

2011-01-01

-electron interactions are described by th=e many-body GW approximation. The conductance follows an exponential length dependence: Gn = Gc exp(-βn). The main difference from standard density functional theory (DFT) calculations is a significant reduction of the contact conductance, Gc, due to an improved alignment......The calculation of the electronic conductance of nanoscale junctions from first principles is a long-standing problem in the field of charge transport. Here we demonstrate excellent agreement with experiments for the transport properties of the gold/alkanediamine benchmark system when electron...

Gyrokinetic Calculations of Microinstabilities and Transport During RF H-Modes on Alcator C-Mod

International Nuclear Information System (INIS)

Redi, M.H.; Fiore, C.; Bonoli, P.; Bourdelle, C.; Budny, R.; Dorland, W.D.; Ernst, D.; Hammett, G.; Mikkelsen, D.; Rice, J.; Wukitch, S.

2002-01-01

Physics understanding for the experimental improvement of particle and energy confinement is being advanced through massively parallel calculations of microturbulence for simulated plasma conditions. The ultimate goal, an experimentally validated, global, non-local, fully nonlinear calculation of plasma microturbulence is still not within reach, but extraordinary progress has been achieved in understanding microturbulence, driving forces and the plasma response in recent years. In this paper we discuss gyrokinetic simulations of plasma turbulence being carried out to examine a reproducible, H-mode, RF heated experiment on the Alcator CMOD tokamak3, which exhibits an internal transport barrier (ITB). This off axis RF case represents the early phase of a very interesting dual frequency RF experiment, which shows density control with central RF heating later in the discharge. The ITB exhibits steep, spontaneous density peaking: a reduction in particle transport occurring without a central particle source. Since the central temperature is maintained while the central density is increasing, this also suggests a thermal transport barrier exists. TRANSP analysis shows that ceff drops inside the ITB. Sawtooth heat pulse analysis also shows a localized thermal transport barrier. For this ICRF EDA H-mode, the minority resonance is at r/a * 0.5 on the high field side. There is a normal shear profile, with q monotonic

CO{sub 2} emissions due to the air transportation in Brazil; Emissoes de CO{sub 2} devido ao transporte aereo no Brasil

Energy Technology Data Exchange (ETDEWEB)

Simoes, Andre Felipe; Schaeffer, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Planejamento Energetico]. E-mail: afsimoes@antares.com.br; roberto@ppe.ufrj.br

2002-07-01

This work intends to to insert and understand the participation of the brazilian air transportation in the ambit of the global climate changes. Firstly an introduction is presented for positioning the Brazil, in the proposed subject; an approach of the tenuous relationship between the air transportation sector and atmospheric environment medium; the energy consumption associated to the growing demand; and the inventory of the CO{sub 2} emissions (Calculated by using the top-down methodology) due to the Brazilian air transportation activities. The work is globally discussed and analysed.

Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

Energy Technology Data Exchange (ETDEWEB)

Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)

2004-03-01

For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)

First-principles calculations of mobility

Science.gov (United States)

Krishnaswamy, Karthik

First-principles calculations can be a powerful predictive tool for studying, modeling and understanding the fundamental scattering mechanisms impacting carrier transport in materials. In the past, calculations have provided important qualitative insights, but numerical accuracy has been limited due to computational challenges. In this talk, we will discuss some of the challenges involved in calculating electron-phonon scattering and carrier mobility, and outline approaches to overcome them. Topics will include the limitations of models for electron-phonon interaction, the importance of grid sampling, and the use of Gaussian smearing to replace energy-conserving delta functions. Using prototypical examples of oxides that are of technological importance-SrTiO3, BaSnO3, Ga2O3, and WO3-we will demonstrate computational approaches to overcome these challenges and improve the accuracy. One approach that leads to a distinct improvement in the accuracy is the use of analytic functions for the band dispersion, which allows for an exact solution of the energy-conserving delta function. For select cases, we also discuss direct quantitative comparisons with experimental results. The computational approaches and methodologies discussed in the talk are general and applicable to other materials, and greatly improve the numerical accuracy of the calculated transport properties, such as carrier mobility, conductivity and Seebeck coefficient. This work was performed in collaboration with B. Himmetoglu, Y. Kang, W. Wang, A. Janotti and C. G. Van de Walle, and supported by the LEAST Center, the ONR EXEDE MURI, and NSF.

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

International Nuclear Information System (INIS)

Matijevic, M.; Grgic, D.; Jecmenica, R.

2016-01-01

This paper presents comparison of the Krsko Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectra with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters (SB cards in SDEF and WWINP file) for MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FW-CADIS methodology. Denovo implements a structured, Cartesian-grid SN solver based on the Koch-Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was first

Computer program for calculation of complex chemical equilibrium compositions and applications. Supplement 1: Transport properties

Science.gov (United States)

Gordon, S.; Mcbride, B.; Zeleznik, F. J.

1984-01-01

An addition to the computer program of NASA SP-273 is given that permits transport property calculations for the gaseous phase. Approximate mixture formulas are used to obtain viscosity and frozen thermal conductivity. Reaction thermal conductivity is obtained by the same method as in NASA TN D-7056. Transport properties for 154 gaseous species were selected for use with the program.

A risk-based sensor placement methodology

International Nuclear Information System (INIS)

Lee, Ronald W.; Kulesz, James J.

2008-01-01

A risk-based sensor placement methodology is proposed to solve the problem of optimal location of sensors to protect population against the exposure to, and effects of, known and/or postulated chemical, biological, and/or radiological threats. Risk is calculated as a quantitative value representing population at risk from exposure at standard exposure levels. Historical meteorological data are used to characterize weather conditions as the frequency of wind speed and direction pairs. The meteorological data drive atmospheric transport and dispersion modeling of the threats, the results of which are used to calculate risk values. Sensor locations are determined via an iterative dynamic programming algorithm whereby threats detected by sensors placed in prior iterations are removed from consideration in subsequent iterations. In addition to the risk-based placement algorithm, the proposed methodology provides a quantification of the marginal utility of each additional sensor. This is the fraction of the total risk accounted for by placement of the sensor. Thus, the criteria for halting the iterative process can be the number of sensors available, a threshold marginal utility value, and/or a minimum cumulative utility achieved with all sensors

Efficient Ab-Initio Electron Transport Calculations for Heterostructures by the Nonequilibrium Green’s Function Method

Directory of Open Access Journals (Sweden)

Hirokazu Takaki

2014-01-01

Full Text Available We present an efficient computation technique for ab-initio electron transport calculations based on density functional theory and the nonequilibrium Green’s function formalism for application to heterostructures with two-dimensional (2D interfaces. The computational load for constructing the Green’s functions, which depends not only on the energy but also on the 2D Bloch wave vector along the interfaces and is thus catastrophically heavy, is circumvented by parallel computational techniques with the message passing interface, which divides the calculations of the Green’s functions with respect to energy and wave vectors. To demonstrate the computational efficiency of the present code, we perform ab-initio electron transport calculations of Al(100-Si(100-Al(100 heterostructures, one of the most typical metal-semiconductor-metal systems, and show their transmission spectra, density of states (DOSs, and dependence on the thickness of the Si layers.

GUIDE TO CALCULATING TRANSPORT EFFICIENCY OF AEROSOLS IN OCCUPATIONAL AIR SAMPLING SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Hogue, M.; Hadlock, D.; Thompson, M.; Farfan, E.

2013-11-12

This report will present hand calculations for transport efficiency based on aspiration efficiency and particle deposition losses. Because the hand calculations become long and tedious, especially for lognormal distributions of aerosols, an R script (R 2011) will be provided for each element examined. Calculations are provided for the most common elements in a remote air sampling system, including a thin-walled probe in ambient air, straight tubing, bends and a sample housing. One popular alternative approach would be to put such calculations in a spreadsheet, a thorough version of which is shared by Paul Baron via the Aerocalc spreadsheet (Baron 2012). To provide greater transparency and to avoid common spreadsheet vulnerabilities to errors (Burns 2012), this report uses R. The particle size is based on the concept of activity median aerodynamic diameter (AMAD). The AMAD is a particle size in an aerosol where fifty percent of the activity in the aerosol is associated with particles of aerodynamic diameter greater than the AMAD. This concept allows for the simplification of transport efficiency calculations where all particles are treated as spheres with the density of water (1g cm-3). In reality, particle densities depend on the actual material involved. Particle geometries can be very complicated. Dynamic shape factors are provided by Hinds (Hinds 1999). Some example factors are: 1.00 for a sphere, 1.08 for a cube, 1.68 for a long cylinder (10 times as long as it is wide), 1.05 to 1.11 for bituminous coal, 1.57 for sand and 1.88 for talc. Revision 1 is made to correct an error in the original version of this report. The particle distributions are based on activity weighting of particles rather than based on the number of particles of each size. Therefore, the mass correction made in the original version is removed from the text and the calculations. Results affected by the change are updated.

MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by the usual Legendre expansion as used in the discrete ordinates codes. Up-scattering may be specified. The geometry is defined by relationships to general 1. or 2. degree surfaces. Array units may be specified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution

A performance assessment methodology for high-level radioactive waste disposal in unsaturated, fractured tuff

International Nuclear Information System (INIS)

Gallegos, D.P.

1991-07-01

Sandia National Laboratories, has developed a methodology for performance assessment of deep geologic disposal of high-level nuclear waste. The applicability of this performance assessment methodology has been demonstrated for disposal in bedded salt and basalt; it has since been modified for assessment of repositories in unsaturated, fractured tuff. Changes to the methodology are primarily in the form of new or modified ground water flow and radionuclide transport codes. A new computer code, DCM3D, has been developed to model three-dimensional ground-water flow in unsaturated, fractured rock using a dual-continuum approach. The NEFTRAN 2 code has been developed to efficiently model radionuclide transport in time-dependent velocity fields, has the ability to use externally calculated pore velocities and saturations, and includes the effect of saturation dependent retardation factors. In order to use these codes together in performance-assessment-type analyses, code-coupler programs were developed to translate DCM3D output into NEFTRAN 2 input. Other portions of the performance assessment methodology were evaluated as part of modifying the methodology for tuff. The scenario methodology developed under the bedded salt program has been applied to tuff. An investigation of the applicability of uncertainty and sensitivity analysis techniques to non-linear models indicate that Monte Carlo simulation remains the most robust technique for these analyses. No changes have been recommended for the dose and health effects models, nor the biosphere transport models. 52 refs., 1 fig

A set of integrated environmental transport and diffusion models for calculating hazardous releases

International Nuclear Information System (INIS)

Pepper, D.W.

1996-01-01

A set of numerical transport and dispersion models is incorporated within a graphical interface shell to predict hazardous material released into the environment. The visual shell (EnviroView) consists of an object-oriented knowledge base, which is used for inventory control, site mapping and orientation, and monitoring of materials. Graphical displays of detailed sites, building locations, floor plans, and three-dimensional views within a room are available to the user using a point and click interface. In the event of a release to the environment, the user can choose from a selection of analytical, finite element, finite volume, and boundary element methods, which calculate atmospheric transport, groundwater transport, and dispersion within a building interior. The program runs on 486 personal computers under WINDOWS

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

Preliminary fee methodology for recovering GTCC-LLW management costs

International Nuclear Information System (INIS)

Clark, L.L.

1990-06-01

The US Department of Energy (DOE) is currently planning a fee to recover costs of managing Greater-Than-Class-C Low-Level Waste (GTCC-LLW). A cash flow basis will be used for fee calculations to ensure recovery of all applicable program costs. Positive cash flows are revenues received from waste generators. Negative cash flows are program expenses for storage, transportation, treatment, and disposal of the wastes and for program development, evaluation, and administration. Program balances are the net result of positive and negative cash flows each year. The methodology calculates fees that will recovery all program expenses taking into account cost inflation. 3 refs., 1 tab

Numerical shoves and countershoves in electron transport calculations

International Nuclear Information System (INIS)

Filippone, W.L.

1986-01-01

The justification for applying the relatively complex (compared to S/sub n/) streaming ray (SR) algorithm to electron transport problems is its potential for doing rapid and accurate calculations. Because of the Lagrangian treatment of the cell-uncollided electrons, the only significant sources of error are the numerical treatment of the scattering kernel and the spatial differencing scheme used for the cell-collided electrons. Considerable progress has been made in reducing the former source of error. If one is willing to pay the price, the latter source of error can be reduced to any desired level by refining the mesh size or by using high-order differencing schemes. Here the method of numerical shoves and countershoves is introduced, which reduces spatial differencing errors using relatively little additional computational effort

Evaluation and comparison of SN and Monte-Carlo charged particle transport calculations

International Nuclear Information System (INIS)

Hadad, K.

2000-01-01

A study was done to evaluate a 3-D S N charged particle transport code called SMARTEPANTS 1 and another 3-D Monte Carlo code called Integrated Tiger Series, ITS 2 . The evaluation study of SMARTEPANTS code was based on angular discretization and reflected boundary sensitivity whilst the evaluation of ITS was based on CPU time and variance reduction. The comparison of the two code was based on energy and charge deposition calculation in block of Gallium Arsenide with embedded gold cylinders. The result of evaluation tests shows that an S 8 calculation maintains both accuracy and speed and calculations with reflected boundaries geometry produces full symmetrical results. As expected for ITS evaluation, the CPU time and variance reduction are opposite to a point beyond which the history augmentation while increasing the CPU time do not result in variance reduction. The comparison test problem showed excellent agreement in total energy deposition calculations

The implementation of the Quality Costs Methodology in the Public Transport Enterprise in Macedonia

Directory of Open Access Journals (Sweden)

Elizabeta Mitreva

2017-02-01

Full Text Available The implementation of TQM (Total Quality Management strategy in the public transport enterprises in Macedonia means improving the quality of services through examination of business processes not just in terms of defining, improvement and design of the process, but also improvement of productivity and optimization of the costs of quality. The purpose of this study is to point out the importance of determining the quality of the transport services, its methods, and techniques for measurement of the optimization of business processes in particular. The analysis of the quality costs when providing transport services can help managers to understand the impact of poor quality on the financial results and the bad image it gives to the enterprise. In this study, we proposed and applied the model for better performance and higher efficiency of the transport enterprise, through the optimization of business processes, change in the corporate culture and use of the complete business potentials. The need for this methodology was imposed as a result of the analysis made in the company in terms of whether is it doing an analysis on the costs of quality or not. The benefits from the utilization of this model will not only lead to increasing the business performance of the transport enterprise, but this model will also serve as a driving force for continuous improvements to the satisfaction of all stakeholders.

Fleet renewal: An approach to achieve sustainable road transport

Directory of Open Access Journals (Sweden)

Manojlović Aleksandar V.

2011-01-01

Full Text Available With more stringent requirements for efficient utilization of energy resources within the transport industry a need for implementation of sustainable development principles has appeared. Such action will be one of competitive advantages in the future. This is especially confirmed within the road transport sector. A methodology implemented in public procurement procedures for fleet renewal regarding the calculation of road vehicles’ operational lifecycle costs has been analyzed in detail in this paper. Afore mentioned calculation comprises the costs for: vehicle ownership, energy, carbon dioxide and pollutants emissions. Implementation of this methodology allows making the choice of energy efficient vehicles and vehicles with notable positive environmental effects. The objective of the research is to assess the influence of specific parameters of vehicle operational lifecycle costs, especially energy costs and estimated vehicle energy consumption, on vehicle choice in the procurement procedure. The case of urban bus fleet in Serbia was analyzed. Their operational lifecycle costs were calculated and differently powered vehicles were assessed. Energy consumption input values were defined. It was proved that defined fleet renewal scenarios could influence unquestionable decrease in energy consumption.

Generalized Bloch Theorem for Complex Periodic Potentials - A Powerful Application to Quantum Transport Calculations

International Nuclear Information System (INIS)

Zhang, Xiaoguang; Varga, Kalman; Pantelides, Sokrates T

2007-01-01

Band-theoretic methods with periodically repeated supercells have been a powerful approach for ground-state electronic structure calculations, but have not so far been adapted for quantum transport problems with open boundary conditions. Here we introduce a generalized Bloch theorem for complex periodic potentials and use a transfer-matrix formulation to cast the transmission probability in a scattering problem with open boundary conditions in terms of the complex wave vectors of a periodic system with absorbing layers, allowing a band technique for quantum transport calculations. The accuracy and utility of the method is demonstrated by the model problems of the transmission of an electron over a square barrier and the scattering of a phonon in an inhomogeneous nanowire. Application to the resistance of a twin boundary in nanocrystalline copper yields excellent agreement with recent experimental data

CLEAR: a model for the calculation of evacuation-time estimates in Emergency Planning Zones

International Nuclear Information System (INIS)

McLean, M.A.; Moeller, M.P.; Desrosiers, A.E.

1983-01-01

This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffice flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffice queues. The program also models the distribution of times required by individuals to prepare for an evacuation. CLEAR can calculate realistic evacuation time estimates using site specific data and can identify troublesome areas within an Emergency Planning Zone

GREET 1.5 - transportation fuel-cycle model - Vol. 1 : methodology, development, use, and results

International Nuclear Information System (INIS)

Wang, M. Q.

1999-01-01

This report documents the development and use of the most recent version (Version 1.5) of the Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model. The model, developed in a spreadsheet format, estimates the full fuel-cycle emissions and energy associated with various transportation fuels and advanced vehicle technologies for light-duty vehicles. The model calculates fuel-cycle emissions of five criteria pollutants (volatile organic compounds, carbon monoxide, nitrogen oxides, particulate matter with diameters of 10 micrometers or less, and sulfur oxides) and three greenhouse gases (carbon dioxide, methane, and nitrous oxide). The model also calculates total energy consumption, fossil fuel consumption, and petroleum consumption when various transportation fuels are used. The GREET model includes the following cycles: petroleum to conventional gasoline, reformulated gasoline, conventional diesel, reformulated diesel, liquefied petroleum gas, and electricity via residual oil; natural gas to compressed natural gas, liquefied natural gas, liquefied petroleum gas, methanol, Fischer-Tropsch diesel, dimethyl ether, hydrogen, and electricity; coal to electricity; uranium to electricity; renewable energy (hydropower, solar energy, and wind) to electricity; corn, woody biomass, and herbaceous biomass to ethanol; soybeans to biodiesel; flared gas to methanol, dimethyl ether, and Fischer-Tropsch diesel; and landfill gases to methanol. This report also presents the results of the analysis of fuel-cycle energy use and emissions associated with alternative transportation fuels and advanced vehicle technologies to be applied to passenger cars and light-duty trucks

Cross sections for electron and photon processes required by electron-transport calculations

International Nuclear Information System (INIS)

Peek, J.M.

1979-11-01

Electron-transport calculations rely on a large collection of electron-atom and photon-atom cross-section data to represent the response characteristics of the target medium. These basic atomic-physics quantities, and certain qualities derived from them that are now commonly in use, are critically reviewed. Publications appearing after 1978 are not given consideration. Processes involving electron or photon energies less than 1 keV are ignored, while an attempt is made to exhaustively cover the remaining independent parameters and target possibilities. Cases for which data improvements can be made from existing information are identified. Ranges of parameters for which state-of-the-art data are not available are sought out, and recommendations for explicit measurements and/or calculations with presently available tools are presented. An attempt is made to identify the maturity of the atomic-physics data and to predict the possibilities for rapid changes in the quality of the data. Finally, weaknesses in the state-of-the-art atomic-physics data and in the conceptual usage of these data in the context of electron-transport theory are discussed. Brief attempts are made to weight the various aspects of these questions and to suggest possible remedies

Dose rate calculations for a reconnaissance vehicle

International Nuclear Information System (INIS)

Grindrod, L.; Mackey, J.; Salmon, M.; Smith, C.; Wall, S.

2005-01-01

A Chemical Nuclear Reconnaissance System (CNRS) has been developed by the British Ministry of Defence to make chemical and radiation measurements on contaminated terrain using appropriate sensors and recording equipment installed in a land rover. A research programme is under way to develop and validate a predictive capability to calculate the build-up of contamination on the vehicle, radiation detector performance and dose rates to the occupants of the vehicle. This paper describes the geometric model of the vehicle and the methodology used for calculations of detector response. Calculated dose rates obtained using the MCBEND Monte Carlo radiation transport computer code in adjoint mode are presented. These address the transient response of the detectors as the vehicle passes through a contaminated area. Calculated dose rates were found to agree with the measured data to be within the experimental uncertainties, thus giving confidence in the shielding model of the vehicle and its application to other scenarios. (authors)

Uneconomical top calculation method

International Nuclear Information System (INIS)

De Noord, M.; Vanm Sambeek, E.J.W.

2003-08-01

The methodology used to calculate the financial gap of renewable electricity sources and technologies is described. This methodology is used for calculating the production subsidy levels (MEP subsidies) for new renewable electricity projects in 2004 and 2005 in the Netherlands [nl

ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

International Nuclear Information System (INIS)

1985-01-01

1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

TRECII: a computer program for transportation risk assessment

International Nuclear Information System (INIS)

Franklin, A.L.

1980-05-01

A risk-based fault tree analysis method has been developed at the Pacific Northwest Laboratory (PNL) for analysis of nuclear fuel cycle operations. This methodology was developed for the Department of Energy (DOE) as a risk analysis tool for evaluating high level waste management systems. A computer package consisting of three programs was written at that time to assist in the performance of risk assessment: ACORN (draws fault trees), MFAULT (analyzes fault trees), and RAFT (calculates risk). This methodology evaluates release consequences and estimates the frequency of occurrence of these consequences. This document describes an additional risk calculating code which can be used in conjunction with two of the three codes for transportation risk assessment. TRECII modifies the definition of risk used in RAFT (prob. x release) to accommodate release consequences in terms of fatalities. Throughout this report risk shall be defined as probability times consequences (fatalities are one possible health effect consequence). This methodology has been applied to a variety of energy material transportation systems. Typically the material shipped has been radioactive, although some adaptation to fossil fuels has occurred. The approach is normally applied to truck or train transport systems with some adaptation to pipelines and aircraft. TRECII is designed to be used primarily in conjunction with MFAULT; however, with a moderate amount of effort by the user, it can be implemented independent of the risk analysis package developed at PNL. Code description and user instructions necessary for the implementation of the TRECII program are provided

Methodology for calculation of doses to man and implementation in Pandora

Energy Technology Data Exchange (ETDEWEB)

Avila, Rodolfo [Facilia AB, Bromma (Sweden); Bergstroem, Ulla [Swepro Project Management AB, Solna (Sweden)

2006-07-15

This report describes methods and data for calculation of doses to man to be used in safety assessments of repositories for nuclear fuel. The methods are based on the latest recommendations from the ICRP; the EU and the national radiation protection authorities. Equations are given for calculation of doses from ingestion of contaminated water and food, inhalation of contaminated air and external exposure from radionuclides in the ground. With the exception of the exposure from food ingestion, the equations are the same used in previous safety assessments. A general equation is suggested for estimation of the exposure from food ingestion, in which the annual demand of carbon is used instead of the annual ingestion of different food-stuffs, which was earlier applied. The report contains tables with recommended values for physiological characteristics such as water intake, food intake and inhalation rates, based on information summarised in an Appendix. Furthermore, tables are given with recommended age dependent dose conversion factors for ingestion and inhalation for a number of nuclides of interest for safety assessments. The most recently published dose conversion factors for external exposure from contaminated ground are also given. An overview of the implementation of the methodology in Pandora, which is the tool that SKB and Posiva currently use for biosphere modelling, is also provided. The work presented in the report is a result from a joint project commissioned by SKB and Posiva.

Methodology for calculation of doses to man and implementation in Pandora

International Nuclear Information System (INIS)

Avila, Rodolfo; Bergstroem, Ulla

2006-07-01

This report describes methods and data for calculation of doses to man to be used in safety assessments of repositories for nuclear fuel. The methods are based on the latest recommendations from the ICRP; the EU and the national radiation protection authorities. Equations are given for calculation of doses from ingestion of contaminated water and food, inhalation of contaminated air and external exposure from radionuclides in the ground. With the exception of the exposure from food ingestion, the equations are the same used in previous safety assessments. A general equation is suggested for estimation of the exposure from food ingestion, in which the annual demand of carbon is used instead of the annual ingestion of different food-stuffs, which was earlier applied. The report contains tables with recommended values for physiological characteristics such as water intake, food intake and inhalation rates, based on information summarised in an Appendix. Furthermore, tables are given with recommended age dependent dose conversion factors for ingestion and inhalation for a number of nuclides of interest for safety assessments. The most recently published dose conversion factors for external exposure from contaminated ground are also given. An overview of the implementation of the methodology in Pandora, which is the tool that SKB and Posiva currently use for biosphere modelling, is also provided. The work presented in the report is a result from a joint project commissioned by SKB and Posiva

The nodal discrete-ordinate transport calculation of anisotropy scattering problem in three-dimensional cartesian geometry

International Nuclear Information System (INIS)

Wu Hongchun; Xie Zhongsheng; Zhu Xuehua

1994-01-01

The nodal discrete-ordinate transport calculating model of anisotropy scattering problem in three-dimensional cartesian geometry is given. The computing code NOTRAN/3D has been encoded and the satisfied conclusion is gained

GRAVITY PIPELINE TRANSPORT FOR HARDENING FILLING MIXTURES

Directory of Open Access Journals (Sweden)

Leonid KROUPNIK

2015-12-01

Full Text Available In underground mining of solid minerals becoming increasingly common development system with stowing hardening mixtures. In this case the natural ore array after it is replaced by an artificial excavation of solidified filling mixture consisting of binder, aggregates and water. Such a mixture is prepared on the surface on special stowing complexes and transported underground at special stowing pipelines. However, it is transported to the horizons of a few kilometers, which requires a sustainable mode of motion of such a mixture in the pipeline. Hardening stowing mixture changes its rheological characteristics over time, which complicates the calculation of the parameters of pipeline transportation. The article suggests a method of determining the initial parameters of such mixtures: the status coefficient, indicator of transportability, coefficient of hydrodynamic resistance to motion of the mixture. These indicators characterize the mixture in terms of the possibility to transport it through pipes. On the basis of these indicators is proposed methodology for calculating the parameters of pipeline transport hardening filling mixtures in drift mode when traffic on the horizontal part of the mixture under pressure column of the mixture in the vertical part of the backfill of the pipeline. This technique allows stable operation is guaranteed to provide pipeline transportation.

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

International Nuclear Information System (INIS)

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-01-01

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

Development of a Quantitative Methodology to Assess the Impacts of Urban Transport Interventions and Related Noise on Well-Being

Directory of Open Access Journals (Sweden)

Matthias Braubach

2015-05-01

Full Text Available Well-being impact assessments of urban interventions are a difficult challenge, as there is no agreed methodology and scarce evidence on the relationship between environmental conditions and well-being. The European Union (EU project “Urban Reduction of Greenhouse Gas Emissions in China and Europe” (URGENCHE explored a methodological approach to assess traffic noise-related well-being impacts of transport interventions in three European cities (Basel, Rotterdam and Thessaloniki linking modeled traffic noise reduction effects with survey data indicating noise-well-being associations. Local noise models showed a reduction of high traffic noise levels in all cities as a result of different urban interventions. Survey data indicated that perception of high noise levels was associated with lower probability of well-being. Connecting the local noise exposure profiles with the noise-well-being associations suggests that the urban transport interventions may have a marginal but positive effect on population well-being. This paper also provides insight into the methodological challenges of well-being assessments and highlights the range of limitations arising from the current lack of reliable evidence on environmental conditions and well-being. Due to these limitations, the results should be interpreted with caution.

Development of a quantitative methodology to assess the impacts of urban transport interventions and related noise on well-being.

Science.gov (United States)

Braubach, Matthias; Tobollik, Myriam; Mudu, Pierpaolo; Hiscock, Rosemary; Chapizanis, Dimitris; Sarigiannis, Denis A; Keuken, Menno; Perez, Laura; Martuzzi, Marco

2015-05-26

Well-being impact assessments of urban interventions are a difficult challenge, as there is no agreed methodology and scarce evidence on the relationship between environmental conditions and well-being. The European Union (EU) project "Urban Reduction of Greenhouse Gas Emissions in China and Europe" (URGENCHE) explored a methodological approach to assess traffic noise-related well-being impacts of transport interventions in three European cities (Basel, Rotterdam and Thessaloniki) linking modeled traffic noise reduction effects with survey data indicating noise-well-being associations. Local noise models showed a reduction of high traffic noise levels in all cities as a result of different urban interventions. Survey data indicated that perception of high noise levels was associated with lower probability of well-being. Connecting the local noise exposure profiles with the noise-well-being associations suggests that the urban transport interventions may have a marginal but positive effect on population well-being. This paper also provides insight into the methodological challenges of well-being assessments and highlights the range of limitations arising from the current lack of reliable evidence on environmental conditions and well-being. Due to these limitations, the results should be interpreted with caution.

Vectorization and parallelization of Monte-Carlo programs for calculation of radiation transport

International Nuclear Information System (INIS)

Seidel, R.

1995-01-01

The versatile MCNP-3B Monte-Carlo code written in FORTRAN77, for simulation of the radiation transport of neutral particles, has been subjected to vectorization and parallelization of essential parts, without touching its versatility. Vectorization is not dependent on a specific computer. Several sample tasks have been selected in order to test the vectorized MCNP-3B code in comparison to the scalar MNCP-3B code. The samples are a representative example of the 3-D calculations to be performed for simulation of radiation transport in neutron and reactor physics. (1) 4πneutron detector. (2) High-energy calorimeter. (3) PROTEUS benchmark (conversion rates and neutron multiplication factors for the HCLWR (High Conversion Light Water Reactor)). (orig./HP) [de

A functional method for estimating DPA tallies in Monte Carlo calculations of Light Water Reactors

International Nuclear Information System (INIS)

Read, Edward A.; Oliveira, Cassiano R.E. de

2011-01-01

There has been a growing need in recent years for the development of methodology to calculate radiation damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is to discuss the development and implementation of a dpa method using Monte Carlo method for transport calculations. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and fuel depletion are assessed for radiation damage calculations and its capability demonstrated and compared to those of the Monte Carlo code MCNP for radiation damage calculations of a typical LWR configuration. (author)

Wind turbines application for energy savings in Gas transportation system

OpenAIRE

Mingaleeva, Renata

2014-01-01

The Thesis shows the perspectives of involving renewable energy resources into the energy balance of Russia, namely the use of wind energy for the purpose of energy supply for the objects of the Russian Gas transportation system. The methodology of the wind energy technical potential calculation is designed and the wind energy technical potential assessment for onshore and offshore zones of Russia is presented. The analysis of Russian Gas transportation system in terms of energy consumption i...

Radionuclide transport calculations from high-level long-lived radioactive waste disposal in deep clayey geologic formation toward adjacent aquifers

International Nuclear Information System (INIS)

Genty, A.; Le Potier, C.

2007-01-01

In the context of high-level nuclear waste repository safety calculations, the modeling of radionuclide migration is of first importance. Three dimensional radionuclide transport calculations in geological repository need to describe objects of the meter scale embedded in geologic layer formations of kilometer extension. A complete and refined spatial description would end up with at least meshes of hundreds of millions to billions elements. The resolution of this kind of problem is today not reachable with classical computers due to resources limitations. Although parallelized computation appears as potential tool to handle multi-scale calculations, to our knowledge no attempt have been yet performed. One emerging solution for repository safety calculations on very large cells meshes consists in using a domain decomposition approach linked to massive parallelized computer calculation. In this approach, the repository domain is divided in small elementary domains and transport calculation are performed independently on different processor for each elementary domain. Before to develop this possible solution, we performed some preliminary test in order to access the order of magnitude of cells needed to perform converged calculation on one elementary disposal domain and to check if Finite Volume (FV) based on Multi Point Flux Approximation (MPFA) spatial scheme or more classical Mixed Hybrid Finite Element (MHFE) spatial scheme were adapted for those calculations in highly heterogeneous porous media. Our preliminary results point out that MHFE and VF schemes applied on non-parallelepiped hexahedral cells for flow and transport calculations in highly heterogeneous media gave satisfactory results. Nevertheless further investigations and additional calculations are needed in order to exhibit the mesh discretization level needed to perform converged calculations. (authors)

3D heterogeneous transport calculations of CANDU fuel with EVENT/HELIOS

International Nuclear Information System (INIS)

Rahnema, F.; Mosher, S.; Ilas, D.; De Oliveira, C.; Eaton, M.; Stamm'ler, R.

2002-01-01

The applicability of the EVENT/HELIOS package to CANDU lattice cell analysis is studied in this paper. A 45-group cross section library is generated using the lattice depletion transport code HELIOS. This library is then used with the 3-D transport code EVENT to compute the pin fission densities and the multiplication constants for six configurations typical of a CANDU cell. The results are compared to those from MCNP with the same multigroup library. Differences of 70-150 pcm in multiplication constant and 0.08-0.95% in pin fission density are found for these cases. It is expected that refining the EVENT calculations can reduce these differences. This gives confidence in applying EVENT to transient analyses at the fuel pin level in a selected part of a CANDU core such as the limiting bundle during a loss of coolant accident (LOCA). (author)

Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.

1988-01-01

The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs

Comparison of Non-overlapping and Overlapping Local/Global Iteration Schemes for Whole-Core Deterministic Transport Calculation

International Nuclear Information System (INIS)

Yuk, Seung Su; Cho, Bumhee; Cho, Nam Zin

2013-01-01

In the case of deterministic transport model, fixed-k problem formulation is necessary and the overlapping local domain is chosen. However, as mentioned in, the partial current-based Coarse Mesh Finite Difference (p-CMFD) procedure enables also non-overlapping local/global (NLG) iteration. In this paper, NLG iteration is combined with p-CMFD and with CMFD (augmented with a concept of p-CMFD), respectively, and compared to OLG iteration on a 2-D test problem. Non-overlapping local/global iteration with p-CMFD and CMFD global calculation is introduced and tested on a 2-D deterministic transport problem. The modified C5G7 problem is analyzed with both NLG and OLG methods and the solutions converge to the reference solution except for some cases of NLG with CMFD. NLG with CMFD gives the best performance if the solution converges. But if fission-source iteration in local calculation is not enough, it is prone to diverge. The p-CMFD global solver gives unconditional convergence (for both OLG and NLG). A study of switching scheme is in progress, where NLG/p-CMFD is used as 'starter' and then switched to NLG/CMFD to render the whole-core transport calculation more efficient and robust. Parallel computation is another obvious future work

Final Report, Nuclear Energy Research Initiative (NERI) Project: An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model

International Nuclear Information System (INIS)

Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hikaru Hiruta; Razvan Nes

2003-01-01

OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations

On calculating phase shifts and performing fits to scattering cross sections or transport properties

International Nuclear Information System (INIS)

Hepburn, J.W.; Roy, R.J. Le

1978-01-01

Improved methods of calculating quantum mechanical phase shifts and for performing least-squares fits to scattering cross sections or transport properties, are described. Their use in a five-parameter fit to experimental differential cross sections reduces the computer time by a factor of 4-7. (Auth.)

Calculation of parameters for an iron shield experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-01-01

In this text is carreid out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gama-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The tranpsort calculations were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reaction and doses rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented reasonable concordance with the experimental measurements. Finally, is presented a proposal for setting up of an experimental arrangement, using the IEA-R1 reactor, with the purpose of lay down a shielding benchmark. (Author) [pt

Calculation and analysis of the mobility and diffusion coefficient of thermal electrons in methane/air premixed flames

KAUST Repository

Bisetti, Fabrizio

2012-12-01

Simulations of ion and electron transport in flames routinely adopt plasma fluid models, which require transport coefficients to compute the mass flux of charged species. In this work, the mobility and diffusion coefficient of thermal electrons in atmospheric premixed methane/air flames are calculated and analyzed. The electron mobility is highest in the unburnt region, decreasing more than threefold across the flame due to mixture composition effects related to the presence of water vapor. Mobility is found to be largely independent of equivalence ratio and approximately equal to 0.4m 2V -1s -1 in the reaction zone and burnt region. The methodology and results presented enable accurate and computationally inexpensive calculations of transport properties of thermal electrons for use in numerical simulations of charged species transport in flames. © 2012 The Combustion Institute.

METHODOLOGY FOR HYDRAULIC CALCULATION OF RIVER REGULATION AND DETERMINATION OF DIKE PARAMETERS

Directory of Open Access Journals (Sweden)

E. I. Mikhnevich

2017-01-01

Full Text Available Territory protection against flood water inundation and creation of polder systems are carried out with the help of protection dikes. One of the main requirements to the composition of polder systems in flood plains is a location of border dikes beyond meander belt in order to avoid their erosion when meander development occurs. Meander belt width can be determined on the basis of the analysis of multi-year land surveying pertaining top river-bed building and in the case when such data is not available this parameter is calculated in accordance with the Snishchenko formula. While banking-up a river bed a flooded area is decreasing and, consequently, water level in inter-dike space and rate of flood water are significantly increasing. For this reason it is necessary to locate dikes at a such distance from a river bed which will not cause rather high increase in water level and flow velocity in the inter-dike space. Methodology for hydraulic calculation of river regulation has been developed in order to substantiate design parameters for levee systems, creation of favourable hydraulic regime in these systems and provision of sustainability for dikes. Its main elements are calculations of pass-through capacity of the leveed channel and rise of water level in inter-dike space, and distance between dikes and their crest level. Peculiar feature of the proposed calculated formulae is an interaction consideration of channel and inundated flows. Their mass-exchanging process results in slowing-down of the channel flow and acceleration of the inundated flow. This occurrence is taken into account and coefficients of kinematic efficiency are introduced to the elements of water flow rate in the river channel and flood plain, respectively. The adduced dependencies for determination of a dike crest level (consequently their height take into consideration a rise of water level in inter-dike space for two types of polder systems: non-inundable (winter dikes with

Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

International Nuclear Information System (INIS)

Apostolov, T; Manolova, M.; Prodanova, R.

2001-01-01

A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion K eff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

Generalized Coarse-Mesh Rebalance Method for Acceleration of Neutron Transport Calculations

International Nuclear Information System (INIS)

Yamamoto, Akio

2005-01-01

This paper proposes a new acceleration method for neutron transport calculations: the generalized coarse-mesh rebalance (GCMR) method. The GCMR method is a unified scheme of the traditional coarse-mesh rebalance (CMR) and the coarse-mesh finite difference (CMFD) acceleration methods. Namely, by using an appropriate acceleration factor, formulation of the GCMR method becomes identical to that of the CMR or CMFD method. This also indicates that the convergence property of the GCMR method can be controlled by the acceleration factor since the convergence properties of the CMR and CMFD methods are generally different. In order to evaluate the convergence property of the GCMR method, a linearized Fourier analysis was carried out for a one-group homogeneous medium, and the results clarified the relationship between the acceleration factor and the spectral radius. It was also shown that the spectral radius of the GCMR method is smaller than those of the CMR and CMFD methods. Furthermore, the Fourier analysis showed that when an appropriate acceleration factor was used, the spectral radius of the GCMR method did not exceed unity in this study, which was in contrast to the results of the CMR or the CMFD method. Application of the GCMR method to practical calculations will be easy when the CMFD acceleration is already adopted in a transport code. By multiplying a suitable acceleration factor to a coefficient (D FD ) of a finite difference formulation, one can improve the numerical instability of the CMFD acceleration method

RADTRAN3, Risk of Radioactive Material Transport

International Nuclear Information System (INIS)

Madsen, M.M.; Taylor, J.M.; Ostmeyer, R.M.; Reardon, P.C.

2001-01-01

1 - Description of program or function: RADTRAN3 is a flexible analytical tool for calculating both the incident-free and accident impacts of transporting radioactive materials. The consequences from incident-free shipments are apportioned among eight population sub- groups and can be calculated for several transport modes. The radiological accident risk (probability times consequence summed over all postulated accidents) is calculated in terms of early fatalities, early morbidities, latent cancer fatalities, genetic effects, and economic impacts. Ground-shine, ingestion, inhalation, direct exposure, resuspension, and cloud-shine dose pathways are modeled to calculate the radiological health risks from accidents. Economic impacts are evaluated based on costs for emergency response, cleanup, evacuation, income loss, and land use. RADTRAN3 can be applied to specific scenario evaluations (individual transport modes or specified combinations), to compare alternative modes or to evaluate generic radioactive material shipments. Unit-risk factors can easily be evaluated to aid in performing generic analyses when several options must be compared with the amount of travel as the only variable. RADTRAN4 offers advances in the handling of route-related data and in the treatment of multiple-isotope materials. 2 - Method of solution: There are several modes used in the transporting of radioactive material such as trucks, passenger vans, passenger airplanes, rail and others. With these modes of transport come several shipment scenarios. The RADTRAN4 methodology uses material, transportation, population distribution, and health effects models to treat the incident-free case. To handle the vehicle accident impacts, accident severity and package release, meteorological dispersion, and economic models are also employed. 3 - Restrictions on the complexity of the problem: There are no apparent limitations due to programming dimensions

A Proposal of Estimation Methodology to Improve Calculation Efficiency of Sampling-based Method in Nuclear Data Sensitivity and Uncertainty Analysis

International Nuclear Information System (INIS)

Song, Myung Sub; Kim, Song Hyun; Kim, Jong Kyung; Noh, Jae Man

2014-01-01

The uncertainty with the sampling-based method is evaluated by repeating transport calculations with a number of cross section data sampled from the covariance uncertainty data. In the transport calculation with the sampling-based method, the transport equation is not modified; therefore, all uncertainties of the responses such as k eff , reaction rates, flux and power distribution can be directly obtained all at one time without code modification. However, a major drawback with the sampling-based method is that it requires expensive computational load for statistically reliable results (inside confidence level 0.95) in the uncertainty analysis. The purpose of this study is to develop a method for improving the computational efficiency and obtaining highly reliable uncertainty result in using the sampling-based method with Monte Carlo simulation. The proposed method is a method to reduce the convergence time of the response uncertainty by using the multiple sets of sampled group cross sections in a single Monte Carlo simulation. The proposed method was verified by estimating GODIVA benchmark problem and the results were compared with that of conventional sampling-based method. In this study, sampling-based method based on central limit theorem is proposed to improve calculation efficiency by reducing the number of repetitive Monte Carlo transport calculation required to obtain reliable uncertainty analysis results. Each set of sampled group cross sections is assigned to each active cycle group in a single Monte Carlo simulation. The criticality uncertainty for the GODIVA problem is evaluated by the proposed and previous method. The results show that the proposed sampling-based method can efficiently decrease the number of Monte Carlo simulation required for evaluate uncertainty of k eff . It is expected that the proposed method will improve computational efficiency of uncertainty analysis with sampling-based method

Analysis and evaluation of critical experiments for validation of neutron transport calculations

International Nuclear Information System (INIS)

Bazzana, S.; Blaumann, H; Marquez Damian, J.I

2009-01-01

The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook. [es

Methodology for calculation of doses to man and implementation in Pandora

International Nuclear Information System (INIS)

Avila, R.; Bergstroem, U.

2006-07-01

This report describes methods and data for calculation of doses to man to be used in safety assessments of repositories for nuclear fuel. The methods are based on the latest recommendations from the ICRP, the EU and the national radiation protection authorities. Equations are given for calculation of doses from ingestion of contaminated water and food, inhalation of contaminated air and external exposure from radionuclides in the ground. With the exception of the exposure from food ingestion, the equations are the same used in previous safety assessments. A general equation is suggested for estimation of the exposure from food ingestion, in which the annual demand of carbon is used instead of the annual ingestion of different foodstuffs, which was earlier applied. The report contains tables with recommended values for physiological characteristics such as water intake, food intake and inhalation rates, based on information summarised in an Appendix. Furthermore, tables are given with recommended age dependent dose conversion factors for ingestion and inhalation for a number of nuclides of interest for safety assessments. The most recently published dose conversion factors for external exposure from contaminated ground are also given. An overview of the implementation of the methodology in Pandora, which is the tool that Posiva and SKB currently use for biosphere modelling, is also provided. The work presented in the report is a result from a joint project commissioned by Svensk Kaernbraenslehantering AB (SKB) and Posiva. The report will be printed also as a SKB report R-06-68. (orig.)

An approximate framework for quantum transport calculation with model order reduction

Energy Technology Data Exchange (ETDEWEB)

Chen, Quan, E-mail: quanchen@eee.hku.hk [Department of Electrical and Electronic Engineering, The University of Hong Kong (Hong Kong); Li, Jun [Department of Chemistry, The University of Hong Kong (Hong Kong); Yam, Chiyung [Beijing Computational Science Research Center (China); Zhang, Yu [Department of Chemistry, The University of Hong Kong (Hong Kong); Wong, Ngai [Department of Electrical and Electronic Engineering, The University of Hong Kong (Hong Kong); Chen, Guanhua [Department of Chemistry, The University of Hong Kong (Hong Kong)

2015-04-01

A new approximate computational framework is proposed for computing the non-equilibrium charge density in the context of the non-equilibrium Green's function (NEGF) method for quantum mechanical transport problems. The framework consists of a new formulation, called the X-formulation, for single-energy density calculation based on the solution of sparse linear systems, and a projection-based nonlinear model order reduction (MOR) approach to address the large number of energy points required for large applied biases. The advantages of the new methods are confirmed by numerical experiments.

Integral transport multiregion geometrical shadowing factor for the approximate collision probability matrix calculation of infinite closely packed lattices

International Nuclear Information System (INIS)

Jowzani-Moghaddam, A.

1981-01-01

An integral transport method of calculating the geometrical shadowing factor in multiregion annular cells for infinite closely packed lattices in cylindrical geometry is developed. This analytical method has been programmed in the TPGS code. This method is based upon a consideration of the properties of the integral transport method for a nonuniform body, which together with Bonalumi's approximations allows the determination of the approximate multiregion collision probability matrix for infinite closely packed lattices with sufficient accuracy. The multiregion geometrical shadowing factors have been calculated for variations in fuel pin annular segment rings in a geometry of annular cells. These shadowing factors can then be used in the calculation of neutron transport from one annulus to another in an infinite lattice. The result of this new geometrical shadowing and collision probability matrix are compared with the Dancoff-Ginsburg correction and the probability matrix using constant shadowing on Yankee fuel elements in an infinite lattice. In these cases the Dancoff-Ginsburg correction factor and collision probability matrix using constant shadowing are in difference by at most 6.2% and 6%, respectively

Calculations of hydrogen transport for the simulation of a Sbo in the NPP-L V using the code CFD GASFLOW

International Nuclear Information System (INIS)

Gomez T, A. M.; Xolocostli M, V.; Lopez M, R.; Filio L, C.; Mugica R, C. A.; Royl, P.

2013-10-01

The scenario of electric power total loss in the nuclear power plant of Laguna Verde (NPP-L V) has been analyzed using the code MELCOR previously, until reaching fault conditions of the primary container. A mitigation measure to avoid the loss of the primary contention is the realization of a venting toward the secondary contention (reactor building), however this measure bears the potential explosions occurrence risk when the hydrogen accumulated in the primary container with the oxygen of the reactor building atmosphere reacting. In this work a scenario has been supposed that considers the mentioned venting when the pressure of 4.5 kg/cm 2 is reached in the primary container. The information for the hydrogen like an entrance fact is obtained of the MELCOR results and the hydrogen transport in both contentions is analyzed with the code CFD GASFLOW that allows predicting the detailed distribution of the hydrogen volumetric concentration and the possible detonation of flammability conditions in the reactor building. The results show that the venting will produce detonation conditions in the venting level (level 33) and flammability in the level of the recharge floor. The methodology here described constitutes the base of a detailed calculation system of this type of phenomena that can use to make safety evaluations in the NPP-L V on scenarios that include gases transport. (Author)

Improving the accuracy of dynamic mass calculation

Directory of Open Access Journals (Sweden)

Oleksandr F. Dashchenko

2015-06-01

Full Text Available With the acceleration of goods transporting, cargo accounting plays an important role in today's global and complex environment. Weight is the most reliable indicator of the materials control. Unlike many other variables that can be measured indirectly, the weight can be measured directly and accurately. Using strain-gauge transducers, weight value can be obtained within a few milliseconds; such values correspond to the momentary load, which acts on the sensor. Determination of the weight of moving transport is only possible by appropriate processing of the sensor signal. The aim of the research is to develop a methodology for weighing freight rolling stock, which increases the accuracy of the measurement of dynamic mass, in particular wagon that moves. Apart from time-series methods, preliminary filtration for improving the accuracy of calculation is used. The results of the simulation are presented.

A SAS2H/KENO-V methodology for 3D fuel burnup analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.

2002-01-01

An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark.44 core. Detailed k sub e sub f sub f and power density evolution with burnup are reported. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Development of a low-level waste risk methodology

International Nuclear Information System (INIS)

Fisher, J.E.; Falconer, K.L.

1984-01-01

A probabilistic risk assessment method is presented for performance evaluation of low-level waste disposal facilities. The associated program package calculates the risk associated with postulated radionuclide release and transport scenarios. Risk is computed as the mathematical product of two statistical variables: the dose consequence of a given release scenario, and its occurrence probability. A sample risk calculation is included which demonstrates the method. This PRA method will facilitate evaluation of facility performance, including identification of high risk scenarios and their mitigation via optimization of site parameters. The method is intended to be used in facility licensing as a demonstration of compliance with the performance objectives set forth in 10 CFR Part 61, or in corresponding state regulations. The Low-Level Waste Risk Methodology is being developed under sponsorship of the Nuclear Regulatory Commission

Aerosol sampling and Transport Efficiency Calculation (ASTEC) and application to surtsey/DCH aerosol sampling system: Code version 1.0: Code description and user's manual

International Nuclear Information System (INIS)

Yamano, N.; Brockmann, J.E.

1989-05-01

This report describes the features and use of the Aerosol Sampling and Transport Efficiency Calculation (ASTEC) Code. The ASTEC code has been developed to assess aerosol transport efficiency source term experiments at Sandia National Laboratories. This code also has broad application for aerosol sampling and transport efficiency calculations in general as well as for aerosol transport considerations in nuclear reactor safety issues. 32 refs., 31 figs., 7 tabs

Model calculation of the characteristic mass for convective and diffusive vapor transport in graphite furnace atomic absorption spectrometry

Energy Technology Data Exchange (ETDEWEB)

Bencs, László, E-mail: bencs.laszlo@wigner.mta.hu [Institute for Solid State Physics and Optics, Wigner Research Centre for Physics, Hungarian Academy of Sciences, P.O. Box 49, H-1525 Budapest (Hungary); Laczai, Nikoletta [Institute for Solid State Physics and Optics, Wigner Research Centre for Physics, Hungarian Academy of Sciences, P.O. Box 49, H-1525 Budapest (Hungary); Ajtony, Zsolt [Institute of Food Science, University of West Hungary, H-9200 Mosonmagyaróvár, Lucsony utca 15–17 (Hungary)

2015-07-01

A combination of former convective–diffusive vapor-transport models is described to extend the calculation scheme for sensitivity (characteristic mass — m{sub 0}) in graphite furnace atomic absorption spectrometry (GFAAS). This approach encompasses the influence of forced convection of the internal furnace gas (mini-flow) combined with concentration diffusion of the analyte atoms on the residence time in a spatially isothermal furnace, i.e., the standard design of the transversely heated graphite atomizer (THGA). A couple of relationships for the diffusional and convectional residence times were studied and compared, including in factors accounting for the effects of the sample/platform dimension and the dosing hole. These model approaches were subsequently applied for the particular cases of Ag, As, Cd, Co, Cr, Cu, Fe, Hg, Mg, Mn, Mo, Ni, Pb, Sb, Se, Sn, V and Zn analytes. For the verification of the accuracy of the calculations, the experimental m{sub 0} values were determined with the application of a standard THGA furnace, operating either under stopped, or mini-flow (50 cm{sup 3} min{sup −1}) of the internal sheath gas during atomization. The theoretical and experimental ratios of m{sub 0}(mini-flow)-to-m{sub 0}(stop-flow) were closely similar for each study analyte. Likewise, the calculated m{sub 0} data gave a fairly good agreement with the corresponding experimental m{sub 0} values for stopped and mini-flow conditions, i.e., it ranged between 0.62 and 1.8 with an average of 1.05 ± 0.27. This indicates the usability of the current model calculations for checking the operation of a given GFAAS instrument and the applied methodology. - Highlights: • A calculation scheme for convective–diffusive vapor loss in GFAAS is described. • Residence time (τ) formulas were compared for sensitivity (m{sub 0}) in a THGA furnace. • Effects of the sample/platform dimension and dosing hole on τ were assessed. • Theoretical m{sub 0} of 18 analytes were

Uncertainty calculation in transport models and forecasts

DEFF Research Database (Denmark)

Manzo, Stefano; Prato, Carlo Giacomo

Transport projects and policy evaluations are often based on transport model output, i.e. traffic flows and derived effects. However, literature has shown that there is often a considerable difference between forecasted and observed traffic flows. This difference causes misallocation of (public...... implemented by using an approach based on stochastic techniques (Monte Carlo simulation and Bootstrap re-sampling) or scenario analysis combined with model sensitivity tests. Two transport models are used as case studies: the Næstved model and the Danish National Transport Model. 3 The first paper...... in a four-stage transport model related to different variable distributions (to be used in a Monte Carlo simulation procedure), assignment procedures and levels of congestion, at both the link and the network level. The analysis used as case study the Næstved model, referring to the Danish town of Næstved2...

Transport appraisal and Monte Carlo simulation by use of the CBA-DK model

DEFF Research Database (Denmark)

Salling, Kim Bang; Leleur, Steen

2011-01-01

calculation, where risk analysis is carried out using Monte Carlo simulation. Special emphasis has been placed on the separation between inherent randomness in the modeling system and lack of knowledge. These two concepts have been defined in terms of variability (ontological uncertainty) and uncertainty......This paper presents the Danish CBA-DK software model for assessment of transport infrastructure projects. The assessment model is based on both a deterministic calculation following the cost-benefit analysis (CBA) methodology in a Danish manual from the Ministry of Transport and on a stochastic...... (epistemic uncertainty). After a short introduction to deterministic calculation resulting in some evaluation criteria a more comprehensive evaluation of the stochastic calculation is made. Especially, the risk analysis part of CBA-DK, with considerations about which probability distributions should be used...

Standardization of the methodology used for fuel pressure drop evaluation to improve hydraulic calculation of heterogeneous cores

International Nuclear Information System (INIS)

Le Borgne, E.; Mattei, A.; Rome, M.; Rodriguez, J.M.

2004-01-01

The determination of hydraulic characteristics for fuel subassembly components is dependent on the hypotheses and the methodology considered. The results of hydraulic compatibility calculations using input data from different sources may thus be difficult to analyse, and their reliability will consequently be reduced. Electricite de France (EDF) and Commissariat a l'Energie Atomique (CEA) have initiated a common program aiming at controlling the consequences of such a situation, increasing the reliability of the values used in the hydraulic compatibility calculations, and proposing a standardization of the operating procedures. In a first step, this program is based on the measurements performed in the CEA HERMES P facility. Extension of this program is expected to the equivalent experimental facilities for which sufficient information will be made available. (author)

A Study on Efficiency Improvement of the Hybrid Monte Carlo/Deterministic Method for Global Transport Problems

International Nuclear Information System (INIS)

Kim, Jong Woo; Woo, Myeong Hyeon; Kim, Jae Hyun; Kim, Do Hyun; Shin, Chang Ho; Kim, Jong Kyung

2017-01-01

In this study hybrid Monte Carlo/Deterministic method is explained for radiation transport analysis in global system. FW-CADIS methodology construct the weight window parameter and it useful at most global MC calculation. However, Due to the assumption that a particle is scored at a tally, less particles are transported to the periphery of mesh tallies. For compensation this space-dependency, we modified the module in the ADVANTG code to add the proposed method. We solved the simple test problem for comparing with result from FW-CADIS methodology, it was confirmed that a uniform statistical error was secured as intended. In the future, it will be added more practical problems. It might be useful to perform radiation transport analysis using the Hybrid Monte Carlo/Deterministic method in global transport problems.

Ab initio calculation of transport properties between PbSe quantum dots facets with iodide ligands

Science.gov (United States)

Wang, B.; Patterson, R.; Chen, W.; Zhang, Z.; Yang, J.; Huang, S.; Shrestha, S.; Conibeer, G.

2018-01-01

The transport properties between Lead Selenide (PbSe) quantum dots decorated with iodide ligands has been studied using density functional theory (DFT). Quantum conductance at each selected energy levels has been calculated along with total density of states and projected density of states. The DFT calculation is carried on using a grid-based planar augmented wave (GPAW) code incorporated with the linear combination of atomic orbital (LCAO) mode and Perdew Burke Ernzerhof (PBE) exchange-correlation functional. Three iodide ligand attached low index facets including (001), (011), (111) are investigated in this work. P-orbital of iodide ligand majorly contributes to density of state (DOS) at near top valence band resulting a significant quantum conductance, whereas DOS of Pb p-orbital shows minor influence. Various values of quantum conductance observed along different planes are possibly reasoned from a combined effect electrical field over topmost surface and total distance between adjacent facets. Ligands attached to (001) and (011) planes possess similar bond length whereas it is significantly shortened in (111) plane, whereas transport between (011) has an overall low value due to newly formed electric field. On the other hand, (111) plane with a net surface dipole perpendicular to surface layers leading to stronger electron coupling suggests an apparent increase of transport probability. Apart from previously mentioned, the maximum transport energy levels located several eVs (1 2 eVs) from the edge of valence band top.

A study of calculation methodology and experimental measurements of the kinetic parameters for source driven subcritical systems

International Nuclear Information System (INIS)

Lee, Seung Min

2009-01-01

This work presents a theoretical study of reactor kinetics focusing on the methodology of calculation and the experimental measurements of the so-called kinetic parameters. A comparison between the methodology based on the Dulla's formalism and the classical method is made. The objective is to exhibit the dependence of the parameters on subcriticality level and perturbation. Two different slab type systems were considered: thermal one and fast one, both with homogeneous media. One group diffusion model was used for the fast reactor, and for the thermal system, two groups diffusion model, considering, in both case, only one precursor's family. The solutions were obtained using the expansion method. Also, descriptions of the main experimental methods of measurements of the kinetic parameters are presented in order to put a question about the compatibility of these methods in subcritical region. (author)

A Methodology for Calculating Radiation Signatures

Energy Technology Data Exchange (ETDEWEB)

Klasky, Marc Louis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilcox, Trevor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bathke, Charles G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); James, Michael R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-05-01

A rigorous formalism is presented for calculating radiation signatures from both Special Nuclear Material (SNM) as well as radiological sources. The use of MCNP6 in conjunction with CINDER/ORIGEN is described to allow for the determination of both neutron and photon leakages from objects of interest. In addition, a description of the use of MCNP6 to properly model the background neutron and photon sources is also presented. Examinations of the physics issues encountered in the modeling are investigated so as to allow for guidance in the user discerning the relevant physics to incorporate into general radiation signature calculations. Furthermore, examples are provided to assist in delineating the pertinent physics that must be accounted for. Finally, examples of detector modeling utilizing MCNP are provided along with a discussion on the generation of Receiver Operating Curves, which are the suggested means by which to determine detectability radiation signatures emanating from objects.

Source-receptor matrix calculation with a Lagrangian particle dispersion model in backward mode

Directory of Open Access Journals (Sweden)

P. Seibert

2004-01-01

Full Text Available The possibility to calculate linear-source receptor relationships for the transport of atmospheric trace substances with a Lagrangian particle dispersion model (LPDM running in backward mode is shown and presented with many tests and examples. This mode requires only minor modifications of the forward LPDM. The derivation includes the action of sources and of any first-order processes (transformation with prescribed rates, dry and wet deposition, radioactive decay, etc.. The backward mode is computationally advantageous if the number of receptors is less than the number of sources considered. The combination of an LPDM with the backward (adjoint methodology is especially attractive for the application to point measurements, which can be handled without artificial numerical diffusion. Practical hints are provided for source-receptor calculations with different settings, both in forward and backward mode. The equivalence of forward and backward calculations is shown in simple tests for release and sampling of particles, pure wet deposition, pure convective redistribution and realistic transport over a short distance. Furthermore, an application example explaining measurements of Cs-137 in Stockholm as transport from areas contaminated heavily in the Chernobyl disaster is included.

Particle Tracking Model and Abstraction of Transport Processes

International Nuclear Information System (INIS)

Robinson, B.

2000-01-01

The purpose of the transport methodology and component analysis is to provide the numerical methods for simulating radionuclide transport and model setup for transport in the unsaturated zone (UZ) site-scale model. The particle-tracking method of simulating radionuclide transport is incorporated into the FEHM computer code and the resulting changes in the FEHM code are to be submitted to the software configuration management system. This Analysis and Model Report (AMR) outlines the assumptions, design, and testing of a model for calculating radionuclide transport in the unsaturated zone at Yucca Mountain. In addition, methods for determining colloid-facilitated transport parameters are outlined for use in the Total System Performance Assessment (TSPA) analyses. Concurrently, process-level flow model calculations are being carrier out in a PMR for the unsaturated zone. The computer code TOUGH2 is being used to generate three-dimensional, dual-permeability flow fields, that are supplied to the Performance Assessment group for subsequent transport simulations. These flow fields are converted to input files compatible with the FEHM code, which for this application simulates radionuclide transport using the particle-tracking algorithm outlined in this AMR. Therefore, this AMR establishes the numerical method and demonstrates the use of the model, but the specific breakthrough curves presented do not necessarily represent the behavior of the Yucca Mountain unsaturated zone

Use of CPXSD for generation of effective fast multigroup libraries for pressure vessel fluence calculations

International Nuclear Information System (INIS)

Alpan, F. Arzu; Haghighat, Alireza

2008-01-01

Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contribution and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine-and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (similar to 1 %-2 %) with the fine-group library from which they were derived, in reaction rate calculations.

Krylov subspace method for evaluating the self-energy matrices in electron transport calculations

DEFF Research Database (Denmark)

Sørensen, Hans Henrik Brandenborg; Hansen, Per Christian; Petersen, D. E.

2008-01-01

We present a Krylov subspace method for evaluating the self-energy matrices used in the Green's function formulation of electron transport in nanoscale devices. A procedure based on the Arnoldi method is employed to obtain solutions of the quadratic eigenvalue problem associated with the infinite...... calculations. Numerical tests within a density functional theory framework are provided to validate the accuracy and robustness of the proposed method, which in most cases is an order of magnitude faster than conventional methods.......We present a Krylov subspace method for evaluating the self-energy matrices used in the Green's function formulation of electron transport in nanoscale devices. A procedure based on the Arnoldi method is employed to obtain solutions of the quadratic eigenvalue problem associated with the infinite...

Comparison of the results of radiation transport calculation obtained by means of different programs

International Nuclear Information System (INIS)

Gorbatkov, D.V.; Kruchkov, V.P.

1995-01-01

Verification of calculational results of radiation transport, obtained by the known, programs and constant libraries (MCNP+ENDF/B, ANISN+HILO, FLUKA92) by means of their comparison with the precision results calculations through ROZ-6N+Sadko program constant complex and with experimental data, is carried out. Satisfactory agreement is shown with the MCNP+ENDF/B package data for the energy range of E<14 MeV. Analysis of the results derivations, obtained trough the ANISN-HILO package for E<400 MeV and the FLUKA92 programs of E<200 GeV is carried out. 25 refs., 12 figs., 3 tabs

AEGIS methodology and a perspective from AEGIS methodology demonstrations

International Nuclear Information System (INIS)

Dove, F.H.

1981-03-01

Objectives of AEGIS (Assessment of Effectiveness of Geologic Isolation Systems) are to develop the capabilities needed to assess the post-closure safety of waste isolation in geologic formation; demonstrate these capabilities on reference sites; apply the assessment methodology to assist the NWTS program in site selection, waste package and repository design; and perform repository site analyses for the licensing needs of NWTS. This paper summarizes the AEGIS methodology, the experience gained from methodology demonstrations, and provides an overview in the following areas: estimation of the response of a repository to perturbing geologic and hydrologic events; estimation of the transport of radionuclides from a repository to man; and assessment of uncertainties

Simplified analytical model for radionuclide transport simulation in the geosphere

International Nuclear Information System (INIS)

Hiromoto, G.

1996-01-01

In order to evaluate postclosure off-site doses from a low-level radioactive waste disposal facilities, an integrated safety assessment methodology has being developed at Instituto de Pesquisas Energeticas e Nucleares. The source-term modelling approach adopted in this system is described and the results obtained in the IAEA NSARS 'The Safety Assessment of Near-Surface Radioactive Waste Disposal Facilities' programme for model intercomparison studies are presented. The radionuclides released from the waste are calculated using a simple first order kinetics model, and the transport, through porous media below the waste is determined by using an analytical solution of the mass transport equation. The methodology and the results obtained in this work are compared with those reported by others participants of the NSARS programme. (author). 4 refs., 4 figs

Real time simulation of the release and transport of radioactive contaminants

International Nuclear Information System (INIS)

Popa, F.; Weber, M.

1991-01-01

Calculating the responses of the radiation monitoring system (RMS) remains one of the most difficult aspects of nuclear power plant simulation to bring into the post-TMI, first principles simulator era. This task requires the simulation of the transport of radioactive contaminants, the transport of the radiation itself, and the instrument channel including the detector. The complex physics and lack of knowledge of input parameters have made these models lag the general simulator trend away from logical/heuristic modeling of physical systems. This paper describes a series of advances to the modeling methodology to change this situation. The objective in the design of this real time simulation model was to always calculate qualitatively reasonable radiation detector readings

A method for determining the spent-fuel contribution to transport cask containment requirements

International Nuclear Information System (INIS)

Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Reardon, P.C.

1992-11-01

This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs

Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

1987-02-01

Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

International Nuclear Information System (INIS)

Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.

2003-01-01

There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

Modeling and simulation of emergent behavior in transportation infrastructure restoration

Science.gov (United States)

Ojha, Akhilesh; Corns, Steven; Shoberg, Thomas G.; Qin, Ruwen; Long, Suzanna K.

2018-01-01

The objective of this chapter is to create a methodology to model the emergent behavior during a disruption in the transportation system and that calculates economic losses due to such a disruption, and to understand how an extreme event affects the road transportation network. The chapter discusses a system dynamics approach which is used to model the transportation road infrastructure system to evaluate the different factors that render road segments inoperable and calculate economic consequences of such inoperability. System dynamics models have been integrated with business process simulation model to evaluate, design, and optimize the business process. The chapter also explains how different factors affect the road capacity. After identifying the various factors affecting the available road capacity, a causal loop diagram (CLD) is created to visually represent the causes leading to a change in the available road capacity and the effects on travel costs when the available road capacity changes.

The modified high-energy transport code, HETC, and design calculations for the SSC [Superconducting Super Collider

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Alsmiller, F.S.; Gabriel, T.A.; Hermann, O.W.; Bishop, B.L.

1988-01-01

The proposed Superconducting Super Collider (SSC) will have two circulating proton beams, each with an energy of 20 TeV. In order to perform detector and shield design calculations at these higher energies that are as accurate as possible, it is necessary to incorporate in the calculations the best available information on differential particle production from hadron-nucleus collisions. In this paper, the manner in which this has been done in the High-Energy Transport Code HETC will be described and calculated results obtained with the modified code will be compared with experimental data. 10 refs., 1 fig

An analytical transport theory method for calculating flux distribution in slab cells

International Nuclear Information System (INIS)

Abdel Krim, M.S.

2001-01-01

A transport theory method for calculating flux distributions in slab fuel cell is described. Two coupled integral equations for flux in fuel and moderator are obtained; assuming partial reflection at moderator external boundaries. Galerkin technique is used to solve these equations. Numerical results for average fluxes in fuel and moderator and the disadvantage factor are given. Comparison with exact numerical methods, that is for total reflection moderator outer boundaries, show that the Galerkin technique gives accurate results for the disadvantage factor and average fluxes. (orig.)

Development of a program for calculating the cells of heavy water

International Nuclear Information System (INIS)

Calabrese, R.; Lerner, A.M.; Notari, C.

1991-01-01

We describe here a methodology to solve the transport equation i cluster-type fuel cells found in PHWR. The general idea is inspired in the English lattice code WIMS-D4 and associated library even if we have introduced innovations both in structure and contents. The different steps involved are the resonant calculation and the subsequent construction of the microscopic self-shielded cross sections for each isotope; the calculation of macroscopic cross sections per material and the condensation to a broader energy structure; the solution of the two dimensional discretized transport equation for the whole cell. The next step is the inclusion of a burn up routine. A program, ALFIN, was written in FORTRAN 77, and prepared in a modular structure. A sample problem is tested and ALFIN results compared to those produce by WIMS-D4. The discrepancies observed are negligible, except for the resonant region where the methods are different and in some aspect WIMS is clearly in error. (author)

CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

International Nuclear Information System (INIS)

Krishnani, P.D.

1992-01-01

The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

Calculation of the poloidal ambipolar field in a stellarator and its effect on transport

International Nuclear Information System (INIS)

Mynick, H.E.

1984-01-01

The portion Phi 1 of the ambipolar potential Phi which produces an electric field in the flux surfaces of a stellarator is self-consistently calculated, and its effect on stellarator transport at low collisionality is considered. The effect is small in a parameter delta/sub h/, which is proportional to the square root of the ripple amplitude, epsilon/sub h/. However, since delta/sub h/ can be an appreciable fraction of 1 for realistic parameters, the effect of Phi 1 on transport can also be appreciable. Whether the effect is harmful or beneficial to confinement depends on the degree of pressure anisotropy and on the sign of p/sub perpendicular/-p/sub parallel/

Evaluation and reffinement of the neutronic calculation methodology

International Nuclear Information System (INIS)

Conti Filho, P.

1984-01-01

A computational code that has the homogenized cross section given by the LEOPARD code as input was developed. The code gives polinomial coefficients that represent the homogenized cross section as a function of the local burnup and the boron concentration for the assembly, for each step in the reactor Burnup. Lately, were developed an interface between the LEOPARD code Polinomiun Generator program and CITATION code to became possible to CITATION code to set the homogenized microscopic cross section as function of the local caracteristics of the assembly on the way to make the calculation of the reactor Burnup. For a choosen reactor (1900MWth) have been done the inicial calculation (super-cells calculation and others Input) and after that were done the calculation with and without the polinomia. The analyses of the results of the CITATION code were done and the principal results were presented here. (Author) [pt

Calculating the Jet Transport Coefficient q-hat in Lattice Gauge Theory

International Nuclear Information System (INIS)

Majumder, Abhijit

2013-01-01

The formalism of jet modification in the higher twist approach is modified to describe a hard parton propagating through a hot thermalized medium. The leading order contribution to the transverse momentum broadening of a high energy (near on-shell) quark in a thermal medium is calculated. This involves a factorization of the perturbative process of scattering of the quark from the non-perturbative transport coefficient. An operator product expansion of the non-perturbative operator product which represents q -hat is carried out and related via dispersion relations to the expectation of local operators. These local operators are then evaluated in quenched SU(2) lattice gauge theory

QmeQ 1.0: An open-source Python package for calculations of transport through quantum dot devices

Science.gov (United States)

Kiršanskas, Gediminas; Pedersen, Jonas Nyvold; Karlström, Olov; Leijnse, Martin; Wacker, Andreas

2017-12-01

QmeQ is an open-source Python package for numerical modeling of transport through quantum dot devices with strong electron-electron interactions using various approximate master equation approaches. The package provides a framework for calculating stationary particle or energy currents driven by differences in chemical potentials or temperatures between the leads which are tunnel coupled to the quantum dots. The electronic structures of the quantum dots are described by their single-particle states and the Coulomb matrix elements between the states. When transport is treated perturbatively to lowest order in the tunneling couplings, the possible approaches are Pauli (classical), first-order Redfield, and first-order von Neumann master equations, and a particular form of the Lindblad equation. When all processes involving two-particle excitations in the leads are of interest, the second-order von Neumann approach can be applied. All these approaches are implemented in QmeQ. We here give an overview of the basic structure of the package, give examples of transport calculations, and outline the range of applicability of the different approximate approaches.

OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

Energy Technology Data Exchange (ETDEWEB)

Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)

2017-06-15

Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

Shifting renewable energy in transport into the next gear. Developing a methodology for taking into account all electricity, hydrogen and methane from renewable sources in the 10% transport target; Hernieuwbare energie in transport naar een hogere versnelling. Ontwikkeling van een methode dat rekening houdt met alle elektriciteit, waterstof en methaan uit hernieuwbare bronnen in de 10% transportdoelsteling

Energy Technology Data Exchange (ETDEWEB)

Kampman, B.; Leguijt, C.; Bennink, D. [CE Delft, Delft (Netherlands); Wentrup, K.; Dreblow, E.; Gruenig, M. [Ecologic Institute, Berlin (Germany); Schmidt, P.; Wurster, R.; Weindorf, W. [Ludwig-Boelkow-Systemtechnik, Muenchen-Ottobrunn (Germany)

2012-01-15

The European Union has set a 10% target of renewable energy use in the transport sector for 2020 in the Renewable Energy Directive (RED, 2009/28/EC). This directive also defines the associated calculation methodologies, for biofuels and renewable electricity used in transport. Regarding biofuels, only those biofuels can contribute that are actually used in the transport sector. The contribution of electricity from renewable sources is treated somewhat differently, as it is typically taken from the electricity grid, where the exact source of the energy used is not monitored: Member States should use the average share of renewable electricity production in their calculations. The RED required the European Commission to present, if appropriate, a proposal to consider the whole amount of the electricity from renewable sources used to power electric vehicles, as well as a methodology to include the contribution of hydrogen from renewable sources in the transport sector. At the same time, there is the question how biomethane injected into the natural gas grid should be counted towards the transport target if vehicles are filled from that same grid - a similar route to that of electricity use in transport. DG Energy of the Commission needs to be supported in the decision making process related to these three routes: renewable electricity, hydrogen and biomethane use in transport, where distribution is taking place via national grids. The result is a comprehensive report in which different methodological options are designed and assessed, and conclusions are drawn, both for the short to medium term (until 2020) and the longer term (post-2020). In the short term, where the contribution of these routes is still limited, a relatively simple approach will be sufficient, but more sophisticated monitoring methodologies may be needed in the future, depending on the way these routes develop [Dutch] In de Richtlijn Hernieuwbare Energie (RED, 2009/28/EC) heeft de Europese Unie

Hanford Site baseline risk assessment methodology

International Nuclear Information System (INIS)

1992-03-01

This report describes risk assessment methodology associated with the remedial action programs at the Hanford Reservation. Topics addressed include human health evaluation, pollutant and radionuclide transport through the environment, and environmental transport pathways

BALTORO a general purpose code for coupling discrete ordinates and Monte-Carlo radiation transport calculations

International Nuclear Information System (INIS)

Zazula, J.M.

1983-01-01

The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)

Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

Energy Technology Data Exchange (ETDEWEB)

Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

2015-07-01

Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

Response surface methodology to simplify calculation of wood energy potency from tropical short rotation coppice species

Science.gov (United States)

Haqiqi, M. T.; Yuliansyah; Suwinarti, W.; Amirta, R.

2018-04-01

Short Rotation Coppice (SRC) system is an option to provide renewable and sustainable feedstock in generating electricity for rural area. Here in this study, we focussed on application of Response Surface Methodology (RSM) to simplify calculation protocols to point out wood chip production and energy potency from some tropical SRC species identified as Bauhinia purpurea, Bridelia tomentosa, Calliandra calothyrsus, Fagraea racemosa, Gliricidia sepium, Melastoma malabathricum, Piper aduncum, Vernonia amygdalina, Vernonia arborea and Vitex pinnata. The result showed that the highest calorific value was obtained from V. pinnata wood (19.97 MJ kg-1) due to its high lignin content (29.84 %, w/w). Our findings also indicated that the use of RSM for estimating energy-electricity of SRC wood had significant term regarding to the quadratic model (R2 = 0.953), whereas the solid-chip ratio prediction was accurate (R2 = 1.000). In the near future, the simple formula will be promising to calculate energy production easily from woody biomass, especially from SRC species.

Quantifying the relative contributions of different solute carriers to aggregate substrate transport

Science.gov (United States)

Taslimifar, Mehdi; Oparija, Lalita; Verrey, Francois; Kurtcuoglu, Vartan; Olgac, Ufuk; Makrides, Victoria

2017-01-01

Determining the contributions of different transporter species to overall cellular transport is fundamental for understanding the physiological regulation of solutes. We calculated the relative activities of Solute Carrier (SLC) transporters using the Michaelis-Menten equation and global fitting to estimate the normalized maximum transport rate for each transporter (Vmax). Data input were the normalized measured uptake of the essential neutral amino acid (AA) L-leucine (Leu) from concentration-dependence assays performed using Xenopus laevis oocytes. Our methodology was verified by calculating Leu and L-phenylalanine (Phe) data in the presence of competitive substrates and/or inhibitors. Among 9 potentially expressed endogenous X. laevis oocyte Leu transporter species, activities of only the uniporters SLC43A2/LAT4 (and/or SLC43A1/LAT3) and the sodium symporter SLC6A19/B0AT1 were required to account for total uptake. Furthermore, Leu and Phe uptake by heterologously expressed human SLC6A14/ATB0,+ and SLC43A2/LAT4 was accurately calculated. This versatile systems biology approach is useful for analyses where the kinetics of each active protein species can be represented by the Hill equation. Furthermore, its applicable even in the absence of protein expression data. It could potentially be applied, for example, to quantify drug transporter activities in target cells to improve specificity. PMID:28091567

Methodology to calculate wall thickness in metallic pipes

International Nuclear Information System (INIS)

Ramirez, G.F.; Feliciano, H.J.

1992-01-01

The principal objective in the developing of the activities of industrial type is to carry out a efficient and productive task: that implies necessarily to know the best working conditions of the equipment and installations to be concerned. The applications of the radioisotope techniques have a long time as useful tools in several fields of human work. For example, in the Petroleos Mexicanos petrochemical complexes, by safety reasons and for to avoid until maximum the losses, it must be know with a high possible precision the operation regimes of the lines of tubes that they conduce the hydrocarbons, with the purpose to know when they should be replaced the defective or wasted pieces. In the Mexican Petroleum Institute is carrying out a work that it has by objective to develop a methodology bases in the use of radioisotopes that permits to determine the average thickness of the metallic tubes wall, that they have thermic insulator, with a precision of ±0.127 mm (±5 thousandth inch). The method is based in the radiation use emitted by Cs-137 sources. In this work it is described the methodology development so as the principal results obtained. (Author)

The Methodology of Selecting the Transport Mode for Companies on the Slovak Transport Market

Directory of Open Access Journals (Sweden)

Černá Lenka

2017-03-01

Full Text Available Transport volume in the Slovak Republic is growing continuously every year. This rising trend is influenced by the development of car industry and its suppliers. Slovak republic has also a geographic strategy position in middle Europe from the side of transport corridors (east-west and north-south. The development of transport volume in freight transport depends on the transport and business processes between the European Union and China and it is an opportunity for Slovak republic to obtain transit transport flows.

Description of a neutron field perturbed by a probe using coupled Monte Carlo and discrete ordinates radiation transport calculations

International Nuclear Information System (INIS)

Zazula, J.M.

1984-01-01

This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs

The discrete cones method for two-dimensional neutron transport calculations

International Nuclear Information System (INIS)

Watanabe, Y.; Maynard, C.W.

1986-01-01

A novel method, the discrete cones method (DC/sub N/), is proposed as an alternative to the discrete ordinates method (S/sub N/) for solutions of the two-dimensional neutron transport equation. The new method utilizes a new concept, discrete cones, which are made by partitioning a unit spherical surface that the direction vector of particles covers. In this method particles in a cone are simultaneously traced instead of those in discrete directions so that an anomaly of the S/sub N/ method, the ray effects, can be eliminated. The DC/sub N/ method has been formulated for X-Y geometry and a program has been creaed by modifying the standard S/sub N/ program TWOTRAN-II. Our sample calculations demonstrate a strong mitigation of the ray effects without a computing cost penalty

Time dependent AN neutron transport calculations in finite media using a numerical inverse Laplace transform technique

International Nuclear Information System (INIS)

Ganapol, B.D.; Sumini, M.

1990-01-01

The time dependent space second order discrete form of the monokinetic transport equation is given an analytical solution, within the Laplace transform domain. Th A n dynamic model is presented and the general resolution procedure is worked out. The solution in the time domain is then obtained through the application of a numerical transform inversion technique. The justification of the research relies in the need to produce reliable and physically meaningful transport benchmarks for dynamic calculations. The paper is concluded by a few results followed by some physical comments

3-D Whole-Core Transport Calculation with 3D/2D Rotational Plane Slicing Method

Energy Technology Data Exchange (ETDEWEB)

Yoo, Han Jong; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

Use of the method of characteristics (MOC) is very popular due to its capability of heterogeneous geometry treatment and widely used for 2-D core calculation, but direct extension of MOC to 3-D core is not so attractive due to huge calculational cost. 2-D/1-D fusion method was very successful for 3-D calculation of current generation reactor types (highly heterogeneous in radial direction but piece-wise homogeneous in axial direction). In this paper, 2-D MOC concept is extended to 3-D core calculation with little modification of an existing 2-D MOC code. The key idea is to suppose 3-D geometry as a set of many 2-D planes like a phone-directory book. Dividing 3-D structure into a large number of 2-D planes and solving each plane with a simple 2-D SN transport method would give the solution of a 3-D structure. This method was developed independently at KAIST but it is found that this concept is similar with that of 'plane tracing' in the MCCG-3D code. The method developed was tested on the 3-D C5G7 OECD/NEA benchmark problem and compared with the 2-D/1-D fusion method. Results show that the proposed method is worth investigating further. A new approach to 3-D whole-core transport calculation is described and tested. By slicing 3-D structure along characteristic planes and solving each 2-D plane problem, we can get 3-D solution. The numerical test results indicate that the new method is comparable with the 2D/1D fusion method and outperforms other existing methods. But more fair comparison should be done in similar discretization level.

Transport theory calculation for a heterogeneous multi-assembly problem by characteristics method with direct neutron path linking technique

International Nuclear Information System (INIS)

Kosaka, Shinya; Saji, Etsuro

2000-01-01

A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted, resulting in a significant reduction in the wall-clock time of the calculation. By all these efforts, coupled with the results of many verification studies, a whole LWR core heterogeneous transport theory calculation finally became practical. CHAPLET is thought to be a useful tool which can produce the reference solutions for analyses of an LWR (author)

Harmonizing carbon footprint calculation for freight transport chains

NARCIS (Netherlands)

Lewis, A.; Ehrler, V.; Auvinen, H.; Maurer, H.; Davydenko, I.; Burmeister, A.; Seidel, S.; Lischke, A.; Kiel, J.

2016-01-01

The European Commission has set as a target a reduction of 60% in transport greenhouse gas emissions by 2050 [EC 11]. This includes freight transport emissions, which present a particular challenge due to the forecast increase in goods transport linked to future economic growth, the current trend of

Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

International Nuclear Information System (INIS)

Whitesides, G.H.; Stephens, M.E.

1984-01-01

During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

A simplified spherical harmonic method for coupled electron-photon transport calculations

International Nuclear Information System (INIS)

Josef, J.A.

1996-12-01

In this thesis we have developed a simplified spherical harmonic method (SP N method) and associated efficient solution techniques for 2-D multigroup electron-photon transport calculations. The SP N method has never before been applied to charged-particle transport. We have performed a first time Fourier analysis of the source iteration scheme and the P 1 diffusion synthetic acceleration (DSA) scheme applied to the 2-D SP N equations. Our theoretical analyses indicate that the source iteration and P 1 DSA schemes are as effective for the 2-D SP N equations as for the 1-D S N equations. Previous analyses have indicated that the P 1 DSA scheme is unstable (with sufficiently forward-peaked scattering and sufficiently small absorption) for the 2-D S N equations, yet is very effective for the 1-D S N equations. In addition, we have applied an angular multigrid acceleration scheme, and computationally demonstrated that it performs as well for the 2-D SP N equations as for the 1-D S N equations. It has previously been shown for 1-D S N calculations that this scheme is much more effective than the DSA scheme when scattering is highly forward-peaked. We have investigated the applicability of the SP N approximation to two different physical classes of problems: satellite electronics shielding from geomagnetically trapped electrons, and electron beam problems. In the space shielding study, the SP N method produced solutions that are accurate within 10% of the benchmark Monte Carlo solutions, and often orders of magnitude faster than Monte Carlo. We have successfully modeled quasi-void problems and have obtained excellent agreement with Monte Carlo. We have observed that the SP N method appears to be too diffusive an approximation for beam problems. This result, however, is in agreement with theoretical expectations

Advanced calculation methodology for manufacturing and technological parameters' uncertainties propagation at arbitrary level of lattice elements grouping

International Nuclear Information System (INIS)

Pecchia, Marco; Vasiliev, Alexander; Leray, Olivier; Ferroukhi, Hakim; Pautz, Andreas

2015-01-01

A new methodology, referred to as manufacturing and technological parameters uncertainty quantification (MTUQ), is under development at Paul Scherrer Institut (PSI). Based on uncertainty and global sensitivity analysis methods, MTUQ aims at advancing state-of-the-art for the treatment of geometrical/material uncertainties in light water reactor computations, using the MCNPX Monte Carlo neutron transport code. The development is currently focused primarily on criticality safety evaluations (CSE). In that context, the key components are a dedicated modular interface with the MCNPX code and a user-friendly interface to model functional relationship between system variables. A unique feature is an automatic capability to parameterize variables belonging to so-called “repeated structures” such as to allow for perturbations of each individual element of a given system modelled with MCNPX. Concerning the statistical analysis capabilities, these are currently implemented through an interface with the ROOT platform to handle the random sampling design. This paper presents the current status of the MTUQ methodology development and a first assessment of an ongoing organisation for economic cooperation and development/nuclear energy agency benchmark dedicated to uncertainty analyses for CSE. The presented results illustrate the overall capabilities of MTUQ and underline its relevance in predicting more realistic results compared to a methodology previously applied at PSI for this particular benchmark. (author)

METHODS OF INTEGRATED OPTIMIZATION MAGLEV TRANSPORT SYSTEMS

Directory of Open Access Journals (Sweden)

A. Lasher

2013-09-01

Full Text Available Purpose. To demonstrate feasibility of the proposed integrated optimization of various MTS parameters to reduce capital investments as well as decrease any operational and maintenance expense. This will make use of MTS reasonable. At present, the Maglev Transport Systems (MTS for High-Speed Ground Transportation (HSGT almost do not apply. Significant capital investments, high operational and maintenance costs are the main reasons why Maglev Transport Systems (MTS are hardly currently used for the High-Speed Ground Transportation (HSGT. Therefore, this article justifies use of Theory of Complex Optimization of Transport (TCOT, developed by one of the co-authors, to reduce MTS costs. Methodology. According to TCOT, authors developed an abstract model of the generalized transport system (AMSTG. This model mathematically determines the optimal balance between all components of the system and thus provides the ultimate adaptation of any transport systems to the conditions of its application. To identify areas for effective use of MTS, by TCOT, the authors developed a dynamic model of distribution and expansion of spheres of effective use of transport systems (DMRRSEPTS. Based on this model, the most efficient transport system was selected for each individual track. The main estimated criterion at determination of efficiency of application of MTS is the size of the specific transportation tariff received from calculation of payback of total given expenses to a standard payback period or term of granting the credit. Findings. The completed multiple calculations of four types of MTS: TRANSRAPID, MLX01, TRANSMAG and TRANSPROGRESS demonstrated efficiency of the integrated optimization of the parameters of such systems. This research made possible expending the scope of effective usage of MTS in about 2 times. The achieved results were presented at many international conferences in Germany, Switzerland, United States, China, Ukraine, etc. Using MTS as an

Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

International Nuclear Information System (INIS)

Oztunali, O.I.; Roles, G.W.

1986-01-01

Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology

Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Oztunali, O.I.; Roles, G.W.

1986-01-01

Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology.

Calculation of the critical buckling of a lattice based on the integral form of the transport equation

International Nuclear Information System (INIS)

Benoist, P.

1990-06-01

The migration area, which relates the buckling to the multiplication factor, can be calculated by means of the Deniz formula. This formula involves the direct and adjoint angular fluxes. It is shown in this note that it is possible, using the integral form of the transport equation, to establish an equivalent formula in which only angle-integrated quantities appear. This formulation is more suitable for the calculation by the collision probably method [fr

A novel method to calculate the extent and amount of drug transported into CSF after intranasal administration.

Science.gov (United States)

Shi, Zhenqi; Zhang, Qizhi; Jiang, Xinguo

2005-01-31

The aim of this paper is to establish a novel method to calculate the extent and amount of drug transported to brain after administration. The cerebrospinal fluid (CSF) was chosen as the target region. The intranasal administration of meptazinol hydrochloride (MEP) was chosen as the model administration and intravenous administration was selected as reference. According to formula transform, the extent was measured by the equation of X(A)CSF, infinity/X0 = Cl(CSF) AUC(0-->infinity)CSF/X0 and the drug amount was calculated by multiplying the dose with the extent. The drug clearance in CSF (Cl(CSF)) was calculated by a method, in which a certain volume of MEP solution was injected directly into rat cistern magna and then clearance was assessed as the reciprocal of the zeroth moment of a CSF level-time curve normalized for dose. In order to testify the accurateness of the method, 14C-sucrose was chosen as reference because of its impermeable characteristic across blood-brain barrier (BBB). It was found out that the MEP concentrations in plasma and CSF after intranasal administration did not show significant difference with those after intravenous administration. However, the extent and amount of MEP transported to CSF was significantly lower compared with those to plasma after these two administrations. In conclusion, the method can be applied to measure the extent and amount of drug transported to CSF, which would be useful to evaluate brain-targeting drug delivery.

Application of an efficient materials perturbation technique to Monte Carlo photon transport calculations in borehole logging

International Nuclear Information System (INIS)

Picton, D.J.; Harris, R.G.; Randle, K.; Weaver, D.R.

1995-01-01

This paper describes a simple, accurate and efficient technique for the calculation of materials perturbation effects in Monte Carlo photon transport calculations. It is particularly suited to the application for which it was developed, namely the modelling of a dual detector density tool as used in borehole logging. However, the method would be appropriate to any photon transport calculation in the energy range 0.1 to 2 MeV, in which the predominant processes are Compton scattering and photoelectric absorption. The method enables a single set of particle histories to provide results for an array of configurations in which material densities or compositions vary. It can calculate the effects of small perturbations very accurately, but is by no means restricted to such cases. For the borehole logging application described here the method has been found to be efficient for a moderate range of variation in the bulk density (of the order of ±30% from a reference value) or even larger changes to a limited portion of the system (e.g. a low density mudcake of the order of a few tens of mm in thickness). The effective speed enhancement over an equivalent set of individual calculations is in the region of an order of magnitude or more. Examples of calculations on a dual detector density tool are given. It is demonstrated that the method predicts, to a high degree of accuracy, the variation of detector count rates with formation density, and that good results are also obtained for the effects of mudcake layers. An interesting feature of the results is that relative count rates (the ratios of count rates obtained with different configurations) can usually be determined more accurately than the absolute values of the count rates. (orig.)

Photon and electron data bases and their use in radiation transport calculations

International Nuclear Information System (INIS)

Cullen, D.E.; Perkins, S.T.; Seltzer, S.M.

1992-02-01

The ENDF/B-VI photon interaction library includes data to describe the interaction of photons with the elements Z=1 to 100 over the energy range 10 eV to 100 MeV. This library has been designed to meet the traditional needs of users to model the interaction and transport of primary photons. However, this library contains additional information which used in a combination with our other data libraries can be used to perform much more detailed calculations, e.g., emission of secondary fluorescence photons. This paper describes both traditional and more detailed uses of this library

A methodology for the estimation of the radiological consequences of a Loss of Coolant Accident

Energy Technology Data Exchange (ETDEWEB)

Kereszturi, Andras; Brolly, Aron; Panka, Istvan; Pazmandi, Tamas; Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). MTA EK, Centre for Energy Research

2017-09-15

For calculation of the radiological consequences of Large Break Loss of Coolant (LBLOCA) events, a set of various computer codes modeling the corresponding physical processes, disciplines and their appropriate subsequent data exchange are necessary. For demonstrating the methodology applied in MTA EK, a LBLOCA event at shut down reactor state - when only limited configuration of the Emergency Core Cooling System (ECCS) is available - was selected. In this special case, fission gas release from a number of fuel pins is obtained from the analyses. This paper describes the initiating event and the corresponding thermal hydraulic calculations and the further physical processes, the necessary models and computer codes and their connections. Additionally the applied conservative assumptions and the Best Estimate Plus Uncertainty (B+U) evaluation applied for characterizing the pin power and burnup distribution in the core are presented. Also, the fuel behavior processes. Finally, the newly developed methodology to predict whether the fuel pins are getting in-hermetic or not is described and the the results of the activity transport and dose calculations are shown.

H2POWER: Development of a methodology to calculate life cycle cost of small and medium-scale hydrogen systems

International Nuclear Information System (INIS)

Verduzco, Laura E.; Duffey, Michael R.; Deason, Jonathan P.

2007-01-01

At this time, hydrogen-based power plants and large hydrogen production facilities are capital intensive and unable to compete financially against hydrocarbon-based energy production facilities. An option to overcome this problem and foster the introduction of hydrogen technology is to introduce small and medium-scale applications such as residential and community hydrogen refueling units. Such units could potentially be used to generate both electricity and heat for the home, as well as hydrogen fuel for the automobile. Cost modeling for the integration of these three forms of energy presents several methodological challenges. This is particularly true since the technology is still in the development phase and both the financial and the environmental cost must be calculated using mainly secondary sources. In order to address these issues and aid in the design of small and medium-scale hydrogen systems, this study presents a computer model to calculate financial and environmental costs of this technology using different hydrogen pathways. The model can design and compare hydrogen refueling units against hydrocarbon-based technologies, including the 'gap' between financial and economic costs. Using the methodology, various penalties and incentives that can foster the introduction of hydrogen-based technologies can be added to the analysis to study their impact on financial cost

Spectral zone selection methodology for pebble bed reactors

International Nuclear Information System (INIS)

Mphahlele, Ramatsemela; Ougouag, Abderrafi M.; Ivanov, Kostadin N.; Gougar, Hans D.

2011-01-01

A methodology is developed for determining boundaries of spectral zones for pebble bed reactors. A spectral zone is defined as a region made up of a number of nodes whose characteristics are collectively similar and that are assigned the same few-group diffusion constants. The spectral zones are selected in such a manner that the difference (error) between the reference transport solution and the diffusion code solution takes a minimum value. This is achieved by choosing spectral zones through optimally minimizing this error. The objective function for the optimization algorithm is the total reaction rate error, which is defined as the sum of the leakage, absorption and fission reaction rates errors in each zone. The selection of these spectral zones is such that the core calculation results based on diffusion theory are within an acceptable tolerance as compared to a proper transport reference solution. Through this work, a consistent approach for identifying spectral zones that yield more accurate diffusion results is introduced.

QmeQ 1.0: An open-source Python package for calculations of transport through quantum dot devices

DEFF Research Database (Denmark)

Kiršanskas, Gediminas; Pedersen, Jonas Nyvold; Karlström, Olov

2017-01-01

QmeQ is an open-source Python package for numerical modeling of transport through quantum dot devices with strong electron–electron interactions using various approximate master equation approaches. The package provides a framework for calculating stationary particle or energy currents driven...

Neoclassical resonant-plateau transport calculation in an effectively axisymmetrized tandem mirror with finite end plate resistance

International Nuclear Information System (INIS)

Katanuma, I.; Kiwamoto, Y.; Adachi, S.; Inutake, M.; Ishii, K.; Yatsu, K.; Sawada, K.; Miyoshi, S.

1987-05-01

Calculations are made for neoclassical resonant-plateau transports in the geometry of the effectively axisymmetrized tandem mirror GAMMA 10 magnetic field, which has minimum B inbord anchors inside the axisymmetric plug/barrier mirror cells. Azimuthal drifts at the local non-axisymmetric regions are included. The radial potential profile is determined by solving selfconsistently the charge neutrality equation. A finite resistance connecting end plate to machine ground provides appropriate boundary conditions on the radial electrostatic potential distribution so that it can be determined uniquely. The calculation is consistent with experimental results of GAMMA 10. (author)

Equilibrium Limit of Boundary Scattering in Carbon Nanostructures: Molecular Dynamics Calculations of Thermal Transport

Science.gov (United States)

Haskins, Justin; Kinaci, Alper; Sevik, Cem; Cagin, Tahir

2012-01-01

It is widely known that graphene and many of its derivative nanostructures have exceedingly high reported thermal conductivities (up to 4000 W/mK at 300 K). Such attractive thermal properties beg the use of these structures in practical devices; however, to implement these materials while preserving transport quality, the influence of structure on thermal conductivity should be thoroughly understood. For graphene nanostructures, having average phonon mean free paths on the order of one micron, a primary concern is how size influences the potential for heat conduction. To investigate this, we employ a novel technique to evaluate the lattice thermal conductivity from the Green-Kubo relations and equilibrium molecular dynamics in systems where phonon-boundary scattering dominates heat flow. Specifically, the thermal conductivities of graphene nanoribbons and carbon nanotubes are calculated in sizes up to 3 microns, and the relative influence of boundary scattering on thermal transport is determined to be dominant at sizes less than 1 micron, after which the thermal transport largely depends on the quality of the nanostructure interface. The method is also extended to carbon nanostructures (fullerenes) where phonon confinement, as opposed to boundary scattering, dominates, and general trends related to the influence of curvature on thermal transport in these materials are discussed.

GRUNCLE, 1. Collision Source Calculation for Program DOT. DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling

International Nuclear Information System (INIS)

1996-01-01

A - Nature of problem or function: DOT solves the Boltzmann transport equation in two-dimensional geometries. Principal applications are to neutron and/or photon transport, although the code can be applied to transport problems for any particles not subject to external force fields. Both homogeneous and external-source problems can be solved. Searches on multiplication factor, time absorption, nuclide concentration, and zone thickness are available for reactor problems. Numerous edits and output data sets for subsequent use are available. DOT-3.5 improves the space-scaling algorithm. DOT-3.5/CAB contains group by group UPSCATTER scaling method. DUCT calculates perturbations to the scalar flux caused by the presence of ducts filled with coolant. VIP is a program for cross section sensitivity analysis using two- dimensional discrete ordinates transport calculations. DGRAD calculates the directional flux gradients from DOT-3 diffusion theory flux tapes. In conjunction with VIP and TPERT, it allows the use of diffusion theory fluxes to obtain exact and first-order perturbation reactivity changes. In order to calculate the reactivity associated with changes in reactor compositions using diffusion theory, it is necessary to fold not only the scalar fluxes with the appropriate cross sections, but also the average flux gradients with the diffusion coefficients. Since DOT diffusion theory does not directly calculate these gradients, it was necessary to calculate the needed quantities external to the DOT code. TPERT is a perturbation code to obtain exact and first-order reactivity changes. TPERT is coupled to VIP which generates adjoint forward flux tables using DOT-3 scalar flux tape information. GRTUNCL calculates an analytical first-collision source for subsequent use in DOT. B - Method of solution: The method of discrete ordinates is used. Balance equations are solved for the density of particles moving along discrete directions in each cell of a two-dimensional spatial

Quantum close coupling calculation of transport and relaxation properties for Hg-H_2 system

International Nuclear Information System (INIS)

Nemati-Kande, Ebrahim; Maghari, Ali

2016-01-01

Highlights: • Several relaxation cross sections are calculated for Hg-H_2 van der Waals complex. • These cross sections are calculated from exact close-coupling method. • Energy-dependent SBE cross sections are calculated for ortho- and para-H_2 + Hg systems. • Viscosity and diffusion coefficients are calculated using Mason-Monchick approximation. • The results obtained by Mason-Monchick approximation are compared to the exact close-coupling results. - Abstract: Quantum mechanical close coupling calculation of the state-to-state transport and relaxation cross sections have been done for Hg-H_2 molecular system using a high-level ab initio potential energy surface. Rotationally averaged cross sections were also calculated to obtain the energy dependent Senftleben-Beenakker cross sections at the energy range of 0.005–25,000 cm"−"1. Boltzmann averaging of the energy dependent Senftleben-Beenakker cross sections showed the temperature dependency over a wide temperature range of 50–2500 K. Interaction viscosity and diffusion coefficients were also calculated using close coupling cross sections and full classical Mason-Monchick approximation. The results were compared with each other and with the available experimental data. It was found that Mason-Monchick approximation for viscosity is more reliable than diffusion coefficient. Furthermore, from the comparison of the experimental diffusion coefficients with the result of the close coupling and Mason-Monchick approximation, it was found that the Hg-H_2 potential energy surface used in this work can reliably predict diffusion coefficient data.

Seismic analysis, support design and stress calculation of HTR-PM transport and conversion devices

International Nuclear Information System (INIS)

Zhang Zheyu; Yuan Chaolong; Zhang Haiquan; Nie Junfeng

2012-01-01

Background: The transport and conversion devices are important guarantees for normal operation of HTR-PM fuel handling system in normal and fault conditions. Purpose: A conflict of devices' support design needs to be solved. The flexibility of supports is required because of pipe thermal expansion displacement, while the stiffness is also required because of large devices quality and eccentric distance. Methods: In this paper, the numerical simulation was employed to analyze the seismic characteristics and optimize the support program, Under the chosen support program, the stress calculation of platen support bracket was designed by solidworks software. Results: The supports solved the conflict between the flexibility and stiffness requirements. Conclusions: Therefore, it can ensure the safety of transport and conversion devices and the supports in seismic conditions. (authors)

Transport and Environment Database System (TRENDS): Maritime Air Pollutant Emission Modelling

DEFF Research Database (Denmark)

Georgakaki, Aliki; Coffey, Robert; Lock, Grahm

2005-01-01

This paper reports the development of the maritime module within the framework of the Transport and Environment Database System (TRENDS) project. A detailed database has been constructed for the calculation of energy consumption and air pollutant emissions. Based on an in-house database...... changes from findings reported in Methodologies for Estimating air pollutant Emissions from Transport (MEET). The database operates on statistical data provided by Eurostat, which describe vessel and freight movements from and towards EU 15 major ports. Data are at port to Maritime Coastal Area (MCA...... with a view to this purpose, are mentioned. Examples of the results obtained by the database are presented. These include detailed air pollutant emission calculations for bulk carriers entering the port of Helsinki, as an example of the database operation, and aggregate results for different types...

Computerization of effluent management and external dose calculation using the 'ODCM' methodology applied to Almaraz-NPP

International Nuclear Information System (INIS)

Garcia Gutierrez, M.E.; Sustacha Duo, D.

1993-01-01

The ODCM (Offsite Dose Calculation Manual), the official operational document for all nuclear power plants develops the details for the technical specifications for discharges and governs their practical application. The use of ODCM methodology for managing and controlling data associated with radioactive discharges, as well as the subsequent processing of this data to assess the radiological impact, requires and generates a large volume of data, which demands the frequent application of laborious and complex calculation processes, making computerization necessary. The computer application created for Almaraz NPP has the capacity to store and manage data on all discharges, evaluate their effects, presents reports and copies the information to be sent periodically to the CSN (Spanish Nuclear Regulatory Commission) on a magnetic tape. The radiological impact of an actual or possible discharge can be evaluated at anytime and, furthermore, general or particular reports and graphs on the discharges and doses over time can be readily obtained. The application is run on a personal computer under a relational database management system. This interactive application is based on menus and windows. (author)

New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

Energy Technology Data Exchange (ETDEWEB)

Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

2016-10-15

The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

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

International Nuclear Information System (INIS)

Hursin, Mathieu; Downar, Thomas J.; Yoon, Joo Il; Joo, Han Gyu

2016-01-01

Highlights: • Reactivity initiated accident analysis with direct whole core transient transport code. • Comparison with usual “two steps” procedure. • Effect of effective delayed neutron fraction definition on energy deposition in the fuel. • Effect of homogenized few-group cross sections generation at the assembly level on energy deposition in the fuel. • Effect of effective fuel temperature definition on energy deposition in the fuel. - Abstract: The impact of the approximations in the “two-steps” procedure used in the current generation of nodal simulators for core transient calculations is assessed by using a higher order solution obtained from a direct, whole core, transient transport calculation. A control rod ejection accident in an idealized minicore is analyzed with PARCS, which uses the two-steps procedure and DeCART which provides the higher order solution. DeCART is used as lattice code to provide the homogenized cross sections and kinetics parameters to PARCS. The approximations made by using (1) the homogenized few-group cross sections and kinetic parameters generated at the assembly level, (2) an effective delayed neutrons fraction, (3) an effective fuel temperature and (4) the few-group formulation are investigated in terms of global and local core power behavior. The results presented in the paper show that the current two-steps procedure produces sufficiently accurate transient results with respect to the direct whole core calculation solution, provided that its parameters are carefully generated using the prescriptions described in the present article.

Diffusion Coefficient Calculations With Low Order Legendre Polynomial and Chebyshev Polynomial Approximation for the Transport Equation in Spherical Geometry

International Nuclear Information System (INIS)

Yasa, F.; Anli, F.; Guengoer, S.

2007-01-01

We present analytical calculations of spherically symmetric radioactive transfer and neutron transport using a hypothesis of P1 and T1 low order polynomial approximation for diffusion coefficient D. Transport equation in spherical geometry is considered as the pseudo slab equation. The validity of polynomial expansionion in transport theory is investigated through a comparison with classic diffusion theory. It is found that for causes when the fluctuation of the scattering cross section dominates, the quantitative difference between the polynomial approximation and diffusion results was physically acceptable in general

Comment on ''Walker diffusion method for calculation of transport properties of composite materials''

International Nuclear Information System (INIS)

Kim, In Chan; Cule, Dinko; Torquato, Salvatore

2000-01-01

In a recent paper [C. DeW. Van Siclen, Phys. Rev. E 59, 2804 (1999)], a random-walk algorithm was proposed as the best method to calculate transport properties of composite materials. It was claimed that the method is applicable both to discrete and continuum systems. The limitations of the proposed algorithm are analyzed. We show that the algorithm does not capture the peculiarities of continuum systems (e.g., ''necks'' or ''choke points'') and we argue that it is the stochastic analog of the finite-difference method. (c) 2000 The American Physical Society

Adaptive Green-Kubo estimates of transport coefficients from molecular dynamics based on robust error analysis

Science.gov (United States)

Jones, Reese E.; Mandadapu, Kranthi K.

2012-04-01

We present a rigorous Green-Kubo methodology for calculating transport coefficients based on on-the-fly estimates of: (a) statistical stationarity of the relevant process, and (b) error in the resulting coefficient. The methodology uses time samples efficiently across an ensemble of parallel replicas to yield accurate estimates, which is particularly useful for estimating the thermal conductivity of semi-conductors near their Debye temperatures where the characteristic decay times of the heat flux correlation functions are large. Employing and extending the error analysis of Zwanzig and Ailawadi [Phys. Rev. 182, 280 (1969)], 10.1103/PhysRev.182.280 and Frenkel [in Proceedings of the International School of Physics "Enrico Fermi", Course LXXV (North-Holland Publishing Company, Amsterdam, 1980)] to the integral of correlation, we are able to provide tight theoretical bounds for the error in the estimate of the transport coefficient. To demonstrate the performance of the method, four test cases of increasing computational cost and complexity are presented: the viscosity of Ar and water, and the thermal conductivity of Si and GaN. In addition to producing accurate estimates of the transport coefficients for these materials, this work demonstrates precise agreement of the computed variances in the estimates of the correlation and the transport coefficient with the extended theory based on the assumption that fluctuations follow a Gaussian process. The proposed algorithm in conjunction with the extended theory enables the calculation of transport coefficients with the Green-Kubo method accurately and efficiently.

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

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Transport calculations with the BALDUR code. Pt. 1

International Nuclear Information System (INIS)

Lackner, K.; Wunderlich, R.

1979-12-01

1-d transport calculations with the BALDUR-code are described for predicting the performance of ZEPHYR under D-T operation. Results presented in this report refer to the impurity-free case, and ion and electron heat conduction losses described by CHIsub(i) = neoclassical and CHIsub(e) = 6.25 x 10 17 /nsub(e) (cgs-units). A simple refuelling scenario taking account of the density limit for the ohmic heating phase, the contribution of neutral injection to the refuelling rate and the need for an approximately balanced D-T mixture at the instance of ignition is adopted. The heating scenario assumes a neutral injection beam with 160 keV particle energy in the main component, with a duration of 1.1 sec. Major radius compression by a factor of 1.5 starts 1 sec after the onset of neutral injection and lasts 100 msec. For this standard scenario the performance is studied in different density regimes and for different neutral injection powers. Under the above assumption ignition is predicted for total neutral injection powers < approx. 16 MW (9.6 MW in the main energy component) and average total β-values < 2.8%. Results including impurities, alternative scaling laws, and deviations from the standard scenario will be presented in another report. (orig.) 891 GG/orig. 892 HIS

A method to assess multi-modal hazmat transport security vulnerabilities: Hazmat transport SVA

NARCIS (Netherlands)

Reniers, G.L.L.; Dullaert, W.E.H.

2013-01-01

The suggested Hazmat transport Security Vulnerability Assessment (SVA) methodology presents a user-friendly approach to determine relative security risk levels of the different modes of hazardous freight transport (i.e., road, railway, inland waterways and pipeline transportation). First, transport

Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

International Nuclear Information System (INIS)

Weinhorst, Bastian; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Wilson, Paul

2015-01-01

Highlights: • Comparison of different approaches for the use of CAD geometry for Monte Carlo transport calculations. • Comparison with regard to user-friendliness and computation performance. • Three approaches, namely conversion with McCad, unstructured mesh feature of MCN6 and DAGMC. • Installation most complex for DAGMC, model preparation worst for McCad, computation performance worst for MCNP6. • Installation easiest for McCad, model preparation best for MCNP6, computation speed fastest for McCad. - Abstract: Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

Energy Technology Data Exchange (ETDEWEB)

Weinhorst, Bastian, E-mail: bastian.weinhorst@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Wilson, Paul [University of Wisconsin-Madison, Computational Nuclear Engineering Research Group, Madison, WI (United States)

2015-10-15

Highlights: • Comparison of different approaches for the use of CAD geometry for Monte Carlo transport calculations. • Comparison with regard to user-friendliness and computation performance. • Three approaches, namely conversion with McCad, unstructured mesh feature of MCN6 and DAGMC. • Installation most complex for DAGMC, model preparation worst for McCad, computation performance worst for MCNP6. • Installation easiest for McCad, model preparation best for MCNP6, computation speed fastest for McCad. - Abstract: Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

Transport survey calculations using the spectral collocation method

International Nuclear Information System (INIS)

Painter, S.L.; Lyon, J.F.

1989-01-01

A novel transport survey code has been developed and is being used to study the sensitivity of stellarator reactor performance to various transport assumptions. Instead of following one of the usual approaches, the steady-state transport equation are solved in integral form using the spectral collocation method. This approach effectively combine the computational efficiency of global models with the general nature of 1-D solutions. A compact torsatron reactor test case was used to study the convergence properties and flexibility of the new method. The heat transport model combined Shaing's model for ripple-induced neoclassical transport, the Chang-Hinton model for axisymmetric neoclassical transport, and neoalcator scaling for anomalous electron heat flux. Alpha particle heating, radiation losses, classical electron-ion heat flow, and external heating were included. For the test problem, the method exhibited some remarkable convergence properties. As the number of basis functions was increased, the maximum, pointwise error in the integrated power balance decayed exponentially until the numerical noise level as reached. Better than 10% accuracy in the globally-averaged quantities was achieved with only 5 basis functions; better than 1% accuracy was achieved with 10 basis functions. The numerical method was also found to be very general. Extreme temperature gradients at the plasma edge which sometimes arise from the neoclassical models and are difficult to resolve with finite-difference methods were easily resolved. 8 refs., 6 figs

DEVELOPMENT OF METHODOLOGY FOR THE CALCULATION OF THE PROJECT INNOVATION INDICATOR AND ITS CRITERIA COMPONENTS

Directory of Open Access Journals (Sweden)

Mariya Vishnevskaya

2017-12-01

Full Text Available Two main components of the problem studied in the article are revealed. At the practical level, the provision of the convenient tools allowing a comprehensive evaluation the proposed innovative project in terms of its possibilities for inclusion in the portfolio or development program, and on the level of science – the need for improvement and complementing the existing methodology of assessment of innovative projects attractiveness in the context of their properties and a specific set of components. The research is scientifically applied since the problem solution involves the science-based development of a set of techniques, allowing the practical use of knowledge gained from large information arrays at the initialization stage. The purpose of the study is the formation of an integrated indicator of the project innovation, with a substantive justification of the calculation method, as a tool for the evaluation and selection of projects to be included in the portfolio of projects and programs. The theoretical and methodological basis of the research is the conceptual provisions and scientific developments of experts on project management issues, published in monographs, periodicals, materials of scientific and practical conferences on the topic of research. The tasks were solved using the general scientific and special methods, mathematical modelling methods based on the system approach. Results. A balanced system of parametric single indicators of innovation is presented – the risks, personnel, quality, innovation, resources, and performers, which allows getting a comprehensive idea of any project already in the initial stages. The choice of a risk tolerance as a key criterion of the “risks” element and the reference characteristics is substantiated, in relation to which it can be argued that the potential project holds promise. A tool for calculating the risk tolerance based on the use of matrices and vector analysis is proposed

Bimodality emerges from transport model calculations of heavy ion collisions at intermediate energy

Science.gov (United States)

Mallik, S.; Das Gupta, S.; Chaudhuri, G.

2016-04-01

This work is a continuation of our effort [S. Mallik, S. Das Gupta, and G. Chaudhuri, Phys. Rev. C 91, 034616 (2015)], 10.1103/PhysRevC.91.034616 to examine if signatures of a phase transition can be extracted from transport model calculations of heavy ion collisions at intermediate energy. A signature of first-order phase transition is the appearance of a bimodal distribution in Pm(k ) in finite systems. Here Pm(k ) is the probability that the maximum of the multiplicity distribution occurs at mass number k . Using a well-known model for event generation [Botzmann-Uehling-Uhlenbeck (BUU) plus fluctuation], we study two cases of central collision: mass 40 on mass 40 and mass 120 on mass 120. Bimodality is seen in both the cases. The results are quite similar to those obtained in statistical model calculations. An intriguing feature is seen. We observe that at the energy where bimodality occurs, other phase-transition-like signatures appear. There are breaks in certain first-order derivatives. We then examine if such breaks appear in standard BUU calculations without fluctuations. They do. The implication is interesting. If first-order phase transition occurs, it may be possible to recognize that from ordinary BUU calculations. Probably the reason this has not been seen already is because this aspect was not investigated before.

Two-group k-eigenvalue benchmark calculations for planar geometry transport in a binary stochastic medium

International Nuclear Information System (INIS)

Davis, I.M.; Palmer, T.S.

2005-01-01

Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water reactor. The mean chord length through the assembly is determined and used as the planar geometry system length. According to assumed or calculated material distributions, this system length is populated with alternating fuel and moderator segments of random size. Neutron flux distributions are numerically computed using a discretized form of the Boltzmann transport equation employing diffusion synthetic acceleration. Average quantities (group fluxes and k-eigenvalue) and variances are calculated from an ensemble of realizations of the mixing statistics. The effects of varying two parameters in the fuel, two different boundary conditions, and three different sets of mixing statistics are assessed. A probability distribution function (PDF) of the k-eigenvalue is generated and compared with previous research. Atomic mix solutions are compared with these benchmark ensemble average flux and k-eigenvalue solutions. Mixing statistics with large standard deviations give the most widely varying ensemble solutions of the flux and k-eigenvalue. The shape of the k-eigenvalue PDF qualitatively agrees with previous work. Its overall shape is independent of variations in fuel cross-sections for the problems considered, but its width is impacted by these variations. Statistical distributions with smaller standard deviations alter the shape of this PDF toward a normal distribution. The atomic mix approximation yields large over-predictions of the ensemble average k-eigenvalue and under-predictions of the flux. Qualitatively correct flux shapes are obtained in some cases. These benchmark calculations indicate that a model which includes higher statistical moments of the mixing statistics is needed for accurate predictions of binary

A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

Energy Technology Data Exchange (ETDEWEB)

Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

2014-08-15

As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

International Nuclear Information System (INIS)

Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.

2014-08-01

As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

Computer codes in nuclear safety, radiation transport and dosimetry

International Nuclear Information System (INIS)

Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.

2006-01-01

The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations

Tritium transport calculations for the IFMIF Tritium Release Test Module

Energy Technology Data Exchange (ETDEWEB)

Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

2014-10-15

Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

Tritium transport calculations for the IFMIF Tritium Release Test Module

International Nuclear Information System (INIS)

Freund, Jana; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

2014-01-01

Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

User's manual for sustainable transportation performance measures calculator

Science.gov (United States)

2010-08-01

Sustainable transportation can be viewed as the provision of safe, effective, and efficient : access and mobility into the future while considering economic, social, and environmental : needs. For the Texas Department of Transportation (TxDOT) to ass...

IRT-type research reactor physical calculation methodology

International Nuclear Information System (INIS)

Carrera, W.; Castaneda, S.; Garcia, F.; Garcia, L.; Reyes, O.

1990-01-01

In the present paper an established physical calculation procedure for the research reactor of the Nuclear Research Center (CIN) is described. The results obtained by the method are compared with the ones reported during the physical start up of a reactor with similar characteristics to the CIN reactor. 11 refs

Environment-based pin-power reconstruction method for homogeneous core calculations

International Nuclear Information System (INIS)

Leroyer, H.; Brosselard, C.; Girardi, E.

2012-01-01

Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

Impurity transport calculations for the limiter shadow region of a tokamak