Site Characterization and Preliminary Performance Assessment Calculation Applied To JAEA-Horonobe URL Site of Japan

Energy Technology Data Exchange (ETDEWEB)

Lim, Doo Hyun [NE Union Hill Road, Suite 200, WA 98052 (United States); Hatanaka, Koichiro; Ishii, Eiichi [Japan Atomic Energy Agency, Hokkaido (Japan)

2010-10-15

JAEA-Horonobe Underground Research Laboratory (URL) is designed for research and development on high-level radioactive waste (HLW) repository in sedimentary rock. For a potential HLW repository, understanding and implementing fracturing and faulting system, with data from the site characterization, into the performance assessment is essential because fracture and fault will be the major conductors or barriers for the groundwater flow and radionuclide release. The objectives are i) quantitative derivation of characteristics and correlation of fracturing/faulting system with geologic and geophysics data obtained from the site characterization, and ii) preliminary performance assessment calculation with characterized site information

Transfer Area Mechanical Handling Calculation

International Nuclear Information System (INIS)

Dianda, B.

2004-01-01

This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAX Company L.L. C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC--28-01R W12101'' (Arthur, W.J., I11 2004). This correspondence was appended by further Correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC--28-OIRW12101; TDL No. 04-024'' (BSC 2004a). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The purpose of this calculation is to establish preliminary bounding equipment envelopes and weights for the Fuel Handling Facility (FHF) transfer areas equipment. This calculation provides preliminary information only to support development of facility layouts and preliminary load calculations. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process. It is intended that this calculation is superseded as the design advances to reflect information necessary to support License Application. The design choices outlined within this calculation represent a demonstration of feasibility and may or may not be included in the completed design. This calculation provides preliminary weight, dimensional envelope, and equipment position in building for the purposes of defining interface variables. This calculation identifies and sizes major equipment and assemblies that dictate overall equipment dimensions and facility interfaces. Sizing of components is based on the selection of commercially available products, where applicable. This is not a specific recommendation for the future use of these components or their

Preliminary Analysis For Wolsong Par Effects Using ISACC Calculations

International Nuclear Information System (INIS)

Song, Yong Mann; Kim, Dong Ha

2012-01-01

In the paper, hydrogen control effects using PARs only are analyzed for severe SBO station blackout (SBO) sequences beyond the design basis accidents in WS-1 which are of CANDU6 type reactor. As a computational tool, the latest version of ISAAC4.3 (Integrated Severe Accident Analysis Code for CANDU), which is a fully integrated and lumped severe accident computer code, is used to simulate hydrogen generation and transport inside the reactor building (R/B) before its failure. For the performance of hydrogen removal, the depletion rate equation of K-PAR developed in Korea is applied. In a CANDU reactor, three areas are identified as sources of hydrogen under severe accidents: fuel-coolant interactions in intact channels, suspended fuel or debris interactions in-calandria tank and debris interactions in-calandria vault. The first two origins provide source for the late ('late' terminology is used because it takes more than one day before calandria tank failure) potential hydrogen combustion before calandria tank failure and all the three origins would provide source for the very late potential hydrogen combustion occurring at or after calaria tank failure. If the hydrogen mitigation system fails, the AICC (adiabatic isochoric complete combustion) burning of highly flammable hydrogen may cause Wolsong R/B failure. So hydrogen induced failure possibility is evaluated, using preliminary ISAAC calculations, under several SBO conditions with and without PAR for both late and very late accident periods

Development and preliminary analyses of material balance evaluation model in nuclear fuel cycle

International Nuclear Information System (INIS)

Matsumura, Tetsuo

1994-01-01

Material balance evaluation model in nuclear fuel cycle has been developed using ORIGEN-2 code as basic engine. This model has feature of: It can treat more than 1000 nuclides including minor actinides and fission products. It has flexibility of modeling and graph output using a engineering work station. I made preliminary calculation of LWR fuel high burnup effect (reloading fuel average burnup of 60 GWd/t) on nuclear fuel cycle. The preliminary calculation shows LWR fuel high burnup has much effect on Japanese Pu balance problem. (author)

A calculation of the ZH → γ H decay in the Littlest Higgs Model

International Nuclear Information System (INIS)

Aranda, J I; Ramirez-Zavaleta, F; Tututi, E S; Cortés-Maldonado, I

2016-01-01

New heavy neutral gauge bosons are predicted in many extensions of the Standard Model, those new bosons are associated with additional gauge symmetries. We present a preliminary calculation of the branching ratio decay for heavy neutral gauge bosons ( Z h ) into γ H in the most popular version of the Little Higgs models. The calculation involves the main contributions at one-loop level induced by fermions, scalars and gauge bosons. Preliminary results show a very suppressed branching ratio of the order of 10 -6 . (paper)

3-D calculations for comparison with the experiments

Energy Technology Data Exchange (ETDEWEB)

Alrsen, A M; Bosser, R

1973-09-27

In order to analyse the axial power profile measurements an attempt has been made to do full 3-D calculations for the Dragon reactor. The calculations are still at a very early stage, but the methods used will be outlined here together with the plans for investigations to be carried out in the near future. Some preliminary-results are reported as no final results have yet been obtained. 3-D calculations are rather expensive because of the computer time consumption. It is therefore essential, before too many big computer jobs are spent, to find approximations which can save calculation time. On the other hand some savings, for instance in the number of mesh points, may cause totally wrong results. The ''proper'' calculations have therefore to be proceeded by a number of preliminary investigations, to ensure optimum accuracy and computer expenses. This report contains some of these preliminary studies.

SU-E-T-632: Preliminary Study On Treating Nose Skin Using Energy and Intensity Modulated Electron Beams with Monte Carlo Based Dose Calculations

International Nuclear Information System (INIS)

Jin, L; Eldib, A; Li, J; Price, R; Ma, C

2015-01-01

Purpose: Uneven nose surfaces and air cavities underneath and the use of bolus present complexity and dose uncertainty when using a single electron energy beam to plan treatments of nose skin with a pencil beam-based planning system. This work demonstrates more accurate dose calculation and more optimal planning using energy and intensity modulated electron radiotherapy (MERT) delivered with a pMLC. Methods: An in-house developed Monte Carlo (MC)-based dose calculation/optimization planning system was employed for treatment planning. Phase space data (6, 9, 12 and 15 MeV) were used as an input source for MC dose calculations for the linac. To reduce the scatter-caused penumbra, a short SSD (61 cm) was used. Our previous work demonstrates good agreement in percentage depth dose and off-axis dose between calculations and film measurement for various field sizes. A MERT plan was generated for treating the nose skin using a patient geometry and a dose volume histogram (DVH) was obtained. The work also shows the comparison of 2D dose distributions between a clinically used conventional single electron energy plan and the MERT plan. Results: The MERT plan resulted in improved target dose coverage as compared to the conventional plan, which demonstrated a target dose deficit at the field edge. The conventional plan showed higher dose normal tissue irradiation underneath the nose skin while the MERT plan resulted in improved conformity and thus reduces normal tissue dose. Conclusion: This preliminary work illustrates that MC-based MERT planning is a promising technique in treating nose skin, not only providing more accurate dose calculation, but also offering an improved target dose coverage and conformity. In addition, this technique may eliminate the necessity of bolus, which often produces dose delivery uncertainty due to the air gaps that may exist between the bolus and skin

FUEL HANDLING FACILITY BACKUP CENTRAL COMMUNICATIONS ROOM SPACE REQUIREMENTS CALCULATION

International Nuclear Information System (INIS)

SZALEWSKI, B.

2005-01-01

The purpose of the Fuel Handling Facility Backup Central Communications Room Space Requirements Calculation is to determine a preliminary estimate of the space required to house the backup central communications room in the Fuel Handling Facility (FHF). This room provides backup communications capability to the primary communication systems located in the Central Control Center Facility. This calculation will help guide FHF designers in allocating adequate space for communications system equipment in the FHF. This is a preliminary calculation determining preliminary estimates based on the assumptions listed in Section 4. As such, there are currently no limitations on the use of this preliminary calculation. The calculations contained in this document were developed by Design and Engineering and are intended solely for the use of Design and Engineering in its work regarding the FHF Backup Central Communications Room Space Requirements. Yucca Mountain Project personnel from Design and Engineering should be consulted before the use of the calculations for purposes other than those stated herein or use by individuals other than authorized personnel in Design and Engineering

A new calculation of atmospheric neutrino flux: the FLUKA approach

International Nuclear Information System (INIS)

Battistoni, G.; Bloise, C.; Cavalli, D.; Ferrari, A.; Montaruli, T.; Rancati, T.; Resconi, S.; Ronga, F.; Sala, P.R.

1999-01-01

Preliminary results from a full 3-D calculation of atmospheric neutrino fluxes using the FLUKA interaction model are presented and compared to previous existing calculations. This effort is motivated mainly by the 3-D capability and the satisfactory degree of accuracy of the hadron-nucleus models embedded in the FLUKA code. Here we show examples of benchmarking tests of the model with cosmic ray experiment results. A comparison of our calculation of the atmospheric neutrino flux with that of the Bartol group, for E ν > 1 GeV, is presented

Many-body calculations with deuteron based single-particle bases and their associated natural orbits

Science.gov (United States)

Puddu, G.

2018-06-01

We use the recently introduced single-particle states obtained from localized deuteron wave-functions as a basis for nuclear many-body calculations. We show that energies can be substantially lowered if the natural orbits (NOs) obtained from this basis are used. We use this modified basis for {}10{{B}}, {}16{{O}} and {}24{{Mg}} employing the bare NNLOopt nucleon–nucleon interaction. The lowering of the energies increases with the mass. Although in principle NOs require a full scale preliminary many-body calculation, we found that an approximate preliminary many-body calculation, with a marginal increase in the computational cost, is sufficient. The use of natural orbits based on an harmonic oscillator basis leads to a much smaller lowering of the energies for a comparable computational cost.

Preliminary safety analysis of the HTTR-IS nuclear hydrogen production system

International Nuclear Information System (INIS)

Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Tachibana, Yukio; Sakaba, Nariaki

2010-06-01

Japan Atomic Energy Agency is planning to demonstrate hydrogen production by thermochemical water-splitting IS process utilizing heat from the high-temperature gas-cooled reactor HTTR (HTTR-IS system). The previous study identified that the HTTR modification due to the coupling of hydrogen production plant requires an additional safety review since the scenario and quantitative values of the evaluation items would be altered from the original HTTR safety review. Hence, preliminary safety analyses are conducted by using the system analysis code. Calculation results showed that evaluation items such as a coolant pressure, temperatures of heat transfer tubes at the pressure boundary, etc., did not exceed allowable values. Also, the peak fuel temperature did not exceed allowable value and therefore the reactor core was not damaged and cooled sufficiently. This report compiles calculation conditions, event scenarios and the calculation results of the preliminary safety analysis. (author)

Entropy Calculation of Reversible Mixing of Ideal Gases Shows Absence of Gibbs Paradox

OpenAIRE

Oleg Borodiouk; Vasili Tatarin

1999-01-01

Abstract: We consider the work of reversible mixing of ideal gases using a real process. Now assumptions were made concerning infinite shifts, infinite number of cycles and infinite work to provide an accurate calculation of entropy resulting from reversible mixing of ideal gases. We derived an equation showing the dependence of this entropy on the difference in potential of mixed gases, which is evidence for the absence of Gibbs' paradox.

Preliminary Calculation for Plasma Chamber Design of Pulsed Electron Source Based on Plasma

International Nuclear Information System (INIS)

Widdi Usada

2009-01-01

This paper described the characteristics of pulsed electron sources with anode-cathode distance of 5 cm, electrode diameter of 10 cm, driven by capacitor energy of 25 J. The preliminary results showed that if the system is operated with diode resistance is 1.6 Ω, plasma resistance is 0.14 Ω, and β is 0.94, the achieved of plasma voltage is 640 V, its current is 4.395 kA with its pulse width of 0.8 μsecond. According to breakdown voltage based on Paschen empirical formula, with this achieved voltage, this system could be operated for operation pressure of 1 torr. (author)

Entropy Calculation of Reversible Mixing of Ideal Gases Shows Absence of Gibbs Paradox

Directory of Open Access Journals (Sweden)

Oleg Borodiouk

1999-05-01

Full Text Available Abstract: We consider the work of reversible mixing of ideal gases using a real process. Now assumptions were made concerning infinite shifts, infinite number of cycles and infinite work to provide an accurate calculation of entropy resulting from reversible mixing of ideal gases. We derived an equation showing the dependence of this entropy on the difference in potential of mixed gases, which is evidence for the absence of Gibbs' paradox.

Preliminary Criticality Calculation on Conceptual Deep Borehole Disposal System for Trans-metal Waste during Operational Phase

International Nuclear Information System (INIS)

Kim, In Young; Choi, Heui Joo; Cho, Dong Geun

2013-01-01

The primary function of any repository is to prevent spreading of dangerous materials into surrounding environment. In the case of high-level radioactive waste repository, radioactive material must be isolated and retarded during sufficient decay time to minimize radiation hazard to human and surrounding environment. Sub-criticality of disposal canister and whole disposal system is minimum requisite to prevent multiplication of radiation hazard. In this study, criticality of disposal canister and DBD system for trans-metal waste is calculated to check compliance of sub-criticality. Preliminary calculation on criticality of conceptual deep borehole disposal system and its canister for trans-metal waste during operational phase is conducted in this study. Calculated criticalities at every temperature are under sub-criticalities and criticalities of canister and DBD system considering temperature are expected to become 0.34932 and 0.37618 approximately. There are obvious limitations in this study. To obtain reliable data, exact elementary composition of each component, system component temperature must be specified and applied, and then proper cross section according to each component temperature must be adopted. However, many assumptions, for example simplified elementary concentration and isothermal component temperature, are adopted in this study. Improvement of these data must be conducted in the future work to progress reliability. And, post closure criticality analyses including geo, thermal, hydro, mechanical, chemical mechanism, especially fissile material re-deposition by precipitation and sorption, must be considered to ascertain criticality safety of DBD system as a future work

Preliminary designs: passive solar manufactured housing. Technical status report

Energy Technology Data Exchange (ETDEWEB)

1980-05-12

The criteria established to guide the development of the preliminary designs are listed. Three preliminary designs incorporating direct gain and/or sunspace are presented. Costs, drawings, and supporting calculations are included. (MHR)

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)

Preliminary test results from a telescope of Hughes pixel arrays at FNAL

International Nuclear Information System (INIS)

Jernigan, J.G.; Arens, J.; Vezie, D.; Collins, T.; Krider, J.; Skubic, P.

1992-09-01

In December of 1991 three silicon hybrid pixel detectors each having 2.56 x 2.56 pixels 30 μm square, made by the Hughes Aircraft Company, were placed in a high energy muon beam at the Fermi National Accelerator Laboratory. Straight tracks were recorded in these detectors at angles to the normal to the plane of the silicon ranging from 0 to 45 degrees. In this note, preliminary results are presented on the straight through tracks, i.e., those passing through the telescope at normal incidence. Pulse height data, signal-to-noise data, and preliminary straight line fits to the data resulting in residual distributions are presented. Preliminary calculations show spatial resolution of less than 5 μm in two dimensions

Magnetic field calculation of the Na-4 muon spectrometer

International Nuclear Information System (INIS)

Cvach, J.; Il'yushchenko, V.I.; Savin, I.A.; Vorozhtsov, S.B.

1980-01-01

A NA-4 muon spectrometer is described. Preliminary results of calculating a magnetic field in a toroidal magnetic detector are given. The spectrometer includes 10 similar supermodules each of which consists of 32 iron discs with 275 cm outer diameter magnetized up to saturation. Each module is an independent detector. The POISSON program is used for calculating magnetic field distribution in a toroidal spectrometer magnet. The results obtained show that a magnetic field of iron is a toroidal one and drops approximately according to the logarithmic law from 21.1 kGs on an inner magnet rig to 17.7 kGs on an outer. Magnet support gives approximately 2 % error

Preliminary considerations concerning actinide solubilities

International Nuclear Information System (INIS)

Newton, T.W.; Bayhurst, B.P.; Daniels, W.R.; Erdal, B.R.; Ogard, A.E.

1980-01-01

Work at the Los Alamos Scientific Laboratory on the fundamental solution chemistry of the actinides has thus far been confined to preliminary considerations of the problems involved in developing an understanding of the precipitation and dissolution behavior of actinide compounds under environmental conditions. Attempts have been made to calculate solubility as a function of Eh and pH using the appropriate thermodynamic data; results have been presented in terms of contour maps showing lines of constant solubility as a function of Eh and pH. Possible methods of control of the redox potential of rock-groundwater systems by the use of Eh buffers (redox couples) is presented

Preliminary shielding calculation for the system of CyberKnife robotic radiosurgery

International Nuclear Information System (INIS)

Toreti, Dalila; Xavier, Clarice; Moura, Fabio

2011-01-01

The CyberKnife robotic system uses a manipulator with six grade of freedom for positioning a 6 MV Linac accelerator for treatment of lesions. This paper presents calculations for a standard room, with 200 cm of thickness walls primary, build for a CyberKnife system, and calculations for a room originally designed for a Linac conventional (with gantry), with secondary barriers of 107 cm thickness. After the realization of shielding for both rooms, the results shown that walls of standard room with 200 cm thickness are adequate for the secondary shield, and for a room with a conventional Linac, from all six evaluated points, two would require additional shielding of nine cm and four cm of concrete with 2.4 g/cubic cm. This shows that the CyberKnife system can be installed in a originally designed room for a conventional Linac with neither restrict nor any shielding, since no incidence of beams on the secondary barriers is existent

Noise exposure in movie theaters: a preliminary study of sound levels during the showing of 25 films.

Science.gov (United States)

Warszawa, Anna; Sataloff, Robert T

2010-09-01

The harmful effects of noise exposure during leisure-time activities are beginning to receive some scrutiny. We conducted a preliminary study to investigate the noise levels during the showings of 25 different films. During each screening, various sound measurements were made with a dosimeter. The movies were classified on the basis of both their Motion Picture Association of America (MPAA) rating and their genre, and the size of the theater and the size of the audience were taken into consideration in the final analysis. Our findings suggest that the sound levels of many movies might be harmful to hearing, although we can draw no definitive conclusions. We did not discern any relationship between noise levels and either MPAA rating or genre. Further studies are recommended.

Preliminary core design calculations for the ACPR Upgrade

International Nuclear Information System (INIS)

Pickard, P.S.

1976-01-01

The goal of the Annular Core Pulse Reactor (ACPR) Upgrade design studies is to define a core configuration that provides a significant increase in pulse fluence and fission energy deposition. The reactor modification should provide as flat an energy deposition profile for experiments as feasible. The fuels examined in this study were UO 2 -BeO (5-15 w/o UO 2 ), UC-ZrC-C (200-500 mg U/cc) and U-ZrH 1.5 . The basic core concept examined was a two region core, - a high heat capacity inner core region surrounded by an outer U-ZrH 1.5 region. Survey core calculations utilizing 1D transport calculations and cross sections libraries derived from the ORNL-AMPX code examined relative fuel loadings, fuel temperatures, reactivity requirements and pulse performance improvement. Reference designs for all candidate fuels were defined utilizing 2D transport and Monte Carlo calculations. The performance implications of alternative core designs were also examined for the UO 2 -BeO and UC-ZrC-C fuel candidates. (author)

The friction of polymers around Tg,Tm : Preliminary results

DEFF Research Database (Denmark)

Sivebæk, Ion Marius; Samoilov, V N; Persson, B N J

We present Molecular Dynamics calculations involving polymers of different lengths. Polymers with lengths from 20 to 1400 carbon atoms are considered. The systems are able to simulate friction between polymer surfaces and polymer against metal. The results we present are very preliminary and they......We present Molecular Dynamics calculations involving polymers of different lengths. Polymers with lengths from 20 to 1400 carbon atoms are considered. The systems are able to simulate friction between polymer surfaces and polymer against metal. The results we present are very preliminary...

Preliminary Calculations of Bypass Flow Distribution in a Multi-Block Air Test

International Nuclear Information System (INIS)

Kim, Min Hwan; Tak, Nam Il

2011-01-01

The development of a methodology for the bypass flow assessment in a prismatic VHTR (Very High Temperature Reactor) core has been conducted at KAERI. A preliminary estimation of variation of local bypass flow gap size between graphite blocks in the NHDD core were carried out. With the predicted gap sizes, their influence on the bypass flow distribution and the core hot spot was assessed. Due to the complexity of gap distributions, a system thermo-fluid analysis code is suggested as a tool for the core thermo-fluid analysis, the model and correlations of which should be validated. In order to generate data for validating the bypass flow analysis model, an experimental facility for a multi-block air test was constructed at Seoul National University (SNU). This study is focused on the preliminary evaluation of flow distribution in the test section to understand how the flow is distributed and to help the selection of experimental case. A commercial CFD code, ANSYS CFX is used for the analyses

Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

International Nuclear Information System (INIS)

Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

2007-12-01

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor

Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

Energy Technology Data Exchange (ETDEWEB)

Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

2007-12-15

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.

Preliminary prediction of inflow into the D-holes at the Stripa Mine

International Nuclear Information System (INIS)

Long, J.C.S.; Karasaki, K.; Davey, A.; Peterson, J.; Landsfeld, M.; Kemeny, J.; Martel, S.

1990-02-01

Lawrence Berkeley Laboratory (LBL) is contracted by the US Department of Energy to provide an auxiliary modeling effort for the Stripa Project. Within this effort, we are making calculations of inflow to the Simulated Drift Experiment (SDE), i.e. inflow to six parallel, closely spaced D-holes, using a preliminary set of data collected in five other holes, the N- and W-holes during Stages 1 and 2 of the Site Characterization and Validation (SCV) project. Our approach has been to focus on the fracture zones rather than the general set of ubiquitous fractures. Approximately 90% of all the water flowing in the rock is flowing in fracture zones which are neither uniformly conductive nor are they infinitely extensive. Our approach has been to adopt the fracture zone locations as they have been identified with geophysics. We use geologic sense and the original geophysical data to add one zone where significant water inflow has been observed that can not be explained with the other geophysical zones. This report covers LBL's preliminary prediction of flow into the D-holes. Care should be taken in interpreting the results given in this report. As explained below, the approach that LBL has designed for developing a fracture hydrology model requires cross-hole hydrologic data. Cross-hole tests are planned for Stage 3 but were unavailable in Stage 1. As such, we have inferred from available data what a cross-hole test might show and used this synthetic data to make a preliminary calculation of the inflow into the D-holes. Then using all the Stage 3 data we will calculate flow into the Validation Drift itself. The report mainly demonstrates the use of our methodology and the simulated results should be considered preliminary

Comparison Between Calculated and Measured Cross Section Changes in Natural Uranium Irradiated in NRX

Energy Technology Data Exchange (ETDEWEB)

Ahlstroem, P E

1961-03-15

It is desirable to obtain an experimental check of the reliability of the methods currently used to determine reactivity changes in a reactor and, with a view to meeting this requirement to some extent, a preliminary comparison has been made between calculated and measured cross-section changes in rods of natural uranium irradiated in NRX. The measurements were made at Harwell in the GLEEP reactor and a description has been given by, inter alia, Ward and Craig. The theory of the calculations, which is briefly described in this report, has been indicated by Littler. The investigation showed that the methods for calculating burn up used at present provides a good illustration of the long-term variations in isotope contents. A satisfactory agreement is obtained with experimental results when calculating apparent cross-section changes in uranium rods due to irradiation if the fission cross- section for {sup 239}Pu is set to 780 b. This is 34 b higher than the figure quoted in BNL - 325 (1958). However, in order to get a good idea as to whether the calculated long-term variations in reactivity really correspond to reality, it is necessary to make further investigations. For this reason the results quoted in this report should be regarded as preliminary.

Biota Modeling in EPA's Preliminary Remediation Goal and Dose Compliance Concentration Calculators for Use in EPA Superfund Risk Assessment: Explanation of Intake Rate Derivation, Transfer Factor Compilation, and Mass Loading Factor Sources

International Nuclear Information System (INIS)

Manning, Karessa L.; Dolislager, Fredrick G.; Bellamy, Michael B.

2016-01-01

The Preliminary Remediation Goal (PRG) and Dose Compliance Concentration (DCC) calculators are screening level tools that set forth Environmental Protection Agency's (EPA) recommended approaches, based upon currently available information with respect to risk assessment, for response actions at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) sites, commonly known as Superfund. The screening levels derived by the PRG and DCC calculators are used to identify isotopes contributing the highest risk and dose as well as establish preliminary remediation goals. Each calculator has a residential gardening scenario and subsistence farmer exposure scenarios that require modeling of the transfer of contaminants from soil and water into various types of biota (crops and animal products). New publications of human intake rates of biota; farm animal intakes of water, soil, and fodder; and soil to plant interactions require updates be implemented into the PRG and DCC exposure scenarios. Recent improvements have been made in the biota modeling for these calculators, including newly derived biota intake rates, more comprehensive soil mass loading factors (MLFs), and more comprehensive soil to tissue transfer factors (TFs) for animals and soil to plant transfer factors (BV's). New biota have been added in both the produce and animal products categories that greatly improve the accuracy and utility of the PRG and DCC calculators and encompass greater geographic diversity on a national and international scale.

Comparison of Calculation Models for Bucket Foundation in Sand

DEFF Research Database (Denmark)

Vaitkunaite, Evelina; Molina, Salvador Devant; Ibsen, Lars Bo

The possibility of fast and rather precise preliminary offshore foundation design is desirable. The ultimate limit state of bucket foundation is investigated using three different geotechnical calculation tools: [Ibsen 2001] an analytical method, LimitState:GEO and Plaxis 3D. The study has focused...... on resultant bearing capacity of variously embedded foundation in sand. The 2D models, [Ibsen 2001] and LimitState:GEO can be used for the preliminary design because they are fast and result in a rather similar bearing capacity calculation compared with the finite element models of Plaxis 3D. The 2D models...

Theory analysis and simple calculation of travelling wave burnup scheme

International Nuclear Information System (INIS)

Zhang Jian; Yu Hong; Gang Zhi

2012-01-01

Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)

Preliminary Analyses Showed Short-Term Mental Health Improvements after a Single-Day Manager Training.

Science.gov (United States)

Boysen, Elena; Schiller, Birgitta; Mörtl, Kathrin; Gündel, Harald; Hölzer, Michael

2018-01-10

Psychosocial working conditions attract more and more attention when it comes to mental health in the workplace. Trying to support managers to deal with their own as well as their employees' psychological risk factors, we conducted a specific manager training. Within this investigation, we wanted to learn about the training's effects and acceptance. A single-day manager training was provided in a large industrial company in Germany. The participants were asked to fill out questionnaires regarding their own physical and mental health condition as well as their working situation. Questionnaires were distributed at baseline, 3-month, and 12-month follow-up. At this point of time the investigation is still ongoing. The current article focuses on short-term preliminary effects. Analyses only included participants that already completed baseline and three months follow-up. Preliminary results from three-month follow-up survey ( n = 33, nmale = 30, Mage = 47.5) indicated positive changes in the manager's mental health condition measured by the Patient Health Questionnaire for depression (PHQ-9: Mt1 = 3.82, Mt2 = 3.15). Training managers about common mental disorders and risk factors at the workplace within a single-day workshop seems to promote positive effects on their own mental health. Especially working with the managers on their own early stress symptoms might have been an important element.

Comparison of calculational methods for EBT reactor nucleonics

International Nuclear Information System (INIS)

Henninger, R.J.; Seed, T.J.; Soran, P.D.; Dudziak, D.J.

1980-01-01

Nucleonic calculations for a preliminary conceptual design of the first wall/blanket/shield/coil assembly for an EBT reactor are described. Two-dimensional Monte Carlo, and one- and two-dimensional discrete-ordinates calculations are compared. Good agreement for the calculated values of tritium breeding and nuclear heating is seen. We find that the three methods are all useful and complementary as a design of this type evolves

Energy mesh optimization for multi-level calculation schemes

International Nuclear Information System (INIS)

Mosca, P.; Taofiki, A.; Bellier, P.; Prevost, A.

2011-01-01

The industrial calculations of third generation nuclear reactors are based on sophisticated strategies of homogenization and collapsing at different spatial and energetic levels. An important issue to ensure the quality of these calculation models is the choice of the collapsing energy mesh. In this work, we show a new approach to generate optimized energy meshes starting from the SHEM 281-group library. The optimization model is applied on 1D cylindrical cells and consists of finding an energy mesh which minimizes the errors between two successive collision probability calculations. The former is realized over the fine SHEM mesh with Livolant-Jeanpierre self-shielded cross sections and the latter is performed with collapsed cross sections over the energy mesh being optimized. The optimization is done by the particle swarm algorithm implemented in the code AEMC and multigroup flux solutions are obtained from standard APOLLO2 solvers. By this new approach, a set of new optimized meshes which encompass from 10 to 50 groups has been defined for PWR and BWR calculations. This set will allow users to adapt the energy detail of the solution to the complexity of the calculation (assembly, multi-assembly, two-dimensional whole core). Some preliminary verifications, in which the accuracy of the new meshes is measured compared to a direct 281-group calculation, show that the 30-group optimized mesh offers a good compromise between simulation time and accuracy for a standard 17 x 17 UO 2 assembly with and without control rods. (author)

SU-C-204-06: Monte Carlo Dose Calculation for Kilovoltage X-Ray-Psoralen Activated Cancer Therapy (X-PACT): Preliminary Results

Energy Technology Data Exchange (ETDEWEB)

Mein, S [Duke University Medical Physics Graduate Program (United States); Gunasingha, R [Department of Radiation Safety, Duke University Medical Center (United States); Nolan, M [Department of Clinical Sciences, College of Veterinary Medicine, North Carolina State University (United States); Oldham, M; Adamson, J [Department of Radiation Oncology, Duke University Medical Center (United States)

2016-06-15

Purpose: X-PACT is an experimental cancer therapy where kV x-rays are used to photo-activate anti-cancer therapeutics through phosphor intermediaries (phosphors that absorb x-rays and re-radiate as UV light). Clinical trials in pet dogs are currently underway (NC State College of Veterinary Medicine) and an essential component is the ability to model the kV dose in these dogs. Here we report the commissioning and characterization of a Monte Carlo (MC) treatment planning simulation tool to calculate X-PACT radiation doses in canine trials. Methods: FLUKA multi-particle MC simulation package was used to simulate a standard X-PACT radiation treatment beam of 80kVp with the Varian OBI x-ray source geometry. The beam quality was verified by comparing measured and simulated attenuation of the beam by various thicknesses of aluminum (2–4.6 mm) under narrow beam conditions (HVL). The beam parameters at commissioning were then corroborated using MC, characterized and verified with empirically collected commissioning data, including: percent depth dose curves (PDD), back-scatter factors (BSF), collimator scatter factor(s), and heel effect, etc. All simulations were conducted for N=30M histories at M=100 iterations. Results: HVL and PDD simulation data agreed with an average percent error of 2.42%±0.33 and 6.03%±1.58, respectively. The mean square error (MSE) values for HVL and PDD (0.07% and 0.50%) were low, as expected; however, longer simulations are required to validate convergence to the expected values. Qualitatively, pre- and post-filtration source spectra matched well with 80kVp references generated via SPEKTR software. Further validation of commissioning data simulation is underway in preparation for first-time 3D dose calculations with canine CBCT data. Conclusion: We have prepared a Monte Carlo simulation capable of accurate dose calculation for use with ongoing X-PACT canine clinical trials. Preliminary results show good agreement with measured data and hold

Biota Modeling in EPA's Preliminary Remediation Goal and Dose Compliance Concentration Calculators for Use in EPA Superfund Risk Assessment: Explanation of Intake Rate Derivation, Transfer Factor Compilation, and Mass Loading Factor Sources

Energy Technology Data Exchange (ETDEWEB)

Manning, Karessa L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dolislager, Fredrick G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bellamy, Michael B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-11-01

The Preliminary Remediation Goal (PRG) and Dose Compliance Concentration (DCC) calculators are screening level tools that set forth Environmental Protection Agency's (EPA) recommended approaches, based upon currently available information with respect to risk assessment, for response actions at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) sites, commonly known as Superfund. The screening levels derived by the PRG and DCC calculators are used to identify isotopes contributing the highest risk and dose as well as establish preliminary remediation goals. Each calculator has a residential gardening scenario and subsistence farmer exposure scenarios that require modeling of the transfer of contaminants from soil and water into various types of biota (crops and animal products). New publications of human intake rates of biota; farm animal intakes of water, soil, and fodder; and soil to plant interactions require updates be implemented into the PRG and DCC exposure scenarios. Recent improvements have been made in the biota modeling for these calculators, including newly derived biota intake rates, more comprehensive soil mass loading factors (MLFs), and more comprehensive soil to tissue transfer factors (TFs) for animals and soil to plant transfer factors (BV's). New biota have been added in both the produce and animal products categories that greatly improve the accuracy and utility of the PRG and DCC calculators and encompass greater geographic diversity on a national and international scale.

Three recent TDHF calculations

International Nuclear Information System (INIS)

Weiss, M.S.

1981-05-01

Three applications of TDHF are discussed. First, vibrational spectra of a post grazing collision 40 Ca nucleus is examined and found to contain many high energy components, qualitatively consistent with recent Orsay experiments. Second, the fusion cross section in energy and angular momentum are calculated for 16 O + 24 Mg to exhibit the parameters of the low l window for this system. A sensitivity of the fusion cross section to the effective two body potential is discussed. Last, a preliminary analysis of 86 Kr + 139 La at E/sub lab/ = 505 MeV calculated in the frozen approximation is displayed, compared to experiment and discussed

Deuteron-induced activation data in EAF for IFMIF calculations

International Nuclear Information System (INIS)

Forrest, R.; Cook, I.

2006-01-01

The main type of activation calculations needed for fusion technology deals with the interaction of neutrons with materials. The road map for development of fusion as an electricity producing technology is based on ITER and IFMIF followed by DEMO. IFMIF is a materials testing facility that will enable materials planned to be used in DEMO to be irradiated to very high fluences, so providing the database of material properties required for the licensing of DEMO. IFMIF will use intense beams of high energy deuterons striking a flowing lithium target to produce the neutron field. Although the neutron spectrum is a good match to those produced in a D-T fusion device, there is a significant high energy tail extending up to 55 MeV. These high energy neutrons were the motivation for increasing the upper energy limit in the neutron-induced part of EAF-2005 so that activation calculations could be made in IFMIF. The deuterons themselves will also make a contribution to activation especially in the target where they strike the lithium but also due to beam losses in the accelerator. It was realised that because of corrosion in the lithium loop there is the potential for a wide range of elements to be present in the target region and it is therefore necessary to have a complete library of deuteron-induced cross section data, just as in the neutron case. A preliminary library based on model calculations with TALYS using global parameters was used to construct a deuteron-induced library and this was released as part of the maintenance release of EAF-2005.1 at the beginning of this year. This data library has been used with an updated version of the inventory code FISPACT to calculate the activation in the lithium target due to reactions of the deuterons with the corrosion products. These calculations show that deuterons are much more important than neutrons (about a factor of 70) in activating the elements other than lithium. This work shows the importance of the effect and means

Preliminary thermal and stress analysis of the SINQ window

International Nuclear Information System (INIS)

Heidenreich, G.

1991-01-01

Preliminary results of a finite element analysis for the SINQ proton beam window are presented. Temperatures and stresses are calculated in an axisymmetric model. As a result of these calculations, the H 2 O-cooled window (safety window) could be redesigned in such a way that plastic deformation resulting from excessive stress in some areas is avoided. (author)

Community males show multiple-perpetrator rape proclivity: development and preliminary validation of an interest scale.

Science.gov (United States)

Alleyne, Emma; Gannon, Theresa A; Ó Ciardha, Caoilte; Wood, Jane L

2014-02-01

The literature on Multiple Perpetrator Rape (MPR) is scant; however, a significant proportion of sexual offending involves multiple perpetrators. In addition to the need for research with apprehended offenders of MPR, there is also a need to conduct research with members of the general public. Recent advances in the forensic literature have led to the development of self-report proclivity scales. These scales have enabled researchers to conduct evaluative studies sampling from members of the general public who may be perpetrators of sexual offenses and have remained undetected, or at highest risk of engaging in sexual offending. The current study describes the development and preliminary validation of the Multiple-Perpetrator Rape Interest Scale (M-PRIS), a vignette-based measure assessing community males' sexual arousal to MPR, behavioral propensity toward MPR and enjoyment of MPR. The findings show that the M-PRIS is a reliable measure of community males' sexual interest in MPR with high internal reliability and temporal stability. In a sample of university males we found that a large proportion (66%) did not emphatically reject an interest in MPR. We also found that rape-supportive cognitive distortions, antisocial attitudes, and high-risk sexual fantasies were predictors of sexual interest in MPR. We discuss these findings and the implications for further research employing proclivity measures referencing theory development and clinical practice.

Poster - 08: Preliminary Investigation into Collapsed-Cone based Dose Calculations for COMS Eye Plaques

International Nuclear Information System (INIS)

Morrison, Hali; Menon, Geetha; Sloboda, Ron

2016-01-01

Purpose: To investigate the accuracy of model-based dose calculations using a collapsed-cone algorithm for COMS eye plaques loaded with I-125 seeds. Methods: The Nucletron SelectSeed 130.002 I-125 seed and the 12 mm COMS eye plaque were incorporated into a research version of the Oncentra® Brachy v4.5 treatment planning system which uses the Advanced Collapsed-cone Engine (ACE) algorithm. Comparisons of TG-43 and high-accuracy ACE doses were performed for a single seed in a 30×30×30 cm 3 water box, as well as with one seed in the central slot of the 12 mm COMS eye plaque. The doses along the plaque central axis (CAX) were used to calculate the carrier correction factor, T(r), and were compared to tabulated and MCNP6 simulated doses for both the SelectSeed and IsoAid IAI-125A seeds. Results: The ACE calculated dose for the single seed in water was on average within 0.62 ± 2.2% of the TG-43 dose, with the largest differences occurring near the end-welds. The ratio of ACE to TG-43 calculated doses along the CAX (T(r)) of the 12 mm COMS plaque for the SelectSeed was on average within 3.0% of previously tabulated data, and within 2.9% of the MCNP6 simulated values. The IsoAid and SelectSeed T(r) values agreed within 0.3%. Conclusions: Initial comparisons show good agreement between ACE and MC doses for a single seed in a 12 mm COMS eye plaque; more complicated scenarios are being investigated to determine the accuracy of this calculation method.

Poster - 08: Preliminary Investigation into Collapsed-Cone based Dose Calculations for COMS Eye Plaques

Energy Technology Data Exchange (ETDEWEB)

Morrison, Hali; Menon, Geetha; Sloboda, Ron [Cross Cancer Institute, Edmonton, AB, and University of Alberta, Edmonton, AB, Cross Cancer Institute, Edmonton, AB, and University of Alberta, Edmonton, AB, Cross Cancer Institute, Edmonton, AB, and University of Alberta, Edmonton, AB (Canada)

2016-08-15

Purpose: To investigate the accuracy of model-based dose calculations using a collapsed-cone algorithm for COMS eye plaques loaded with I-125 seeds. Methods: The Nucletron SelectSeed 130.002 I-125 seed and the 12 mm COMS eye plaque were incorporated into a research version of the Oncentra® Brachy v4.5 treatment planning system which uses the Advanced Collapsed-cone Engine (ACE) algorithm. Comparisons of TG-43 and high-accuracy ACE doses were performed for a single seed in a 30×30×30 cm{sup 3} water box, as well as with one seed in the central slot of the 12 mm COMS eye plaque. The doses along the plaque central axis (CAX) were used to calculate the carrier correction factor, T(r), and were compared to tabulated and MCNP6 simulated doses for both the SelectSeed and IsoAid IAI-125A seeds. Results: The ACE calculated dose for the single seed in water was on average within 0.62 ± 2.2% of the TG-43 dose, with the largest differences occurring near the end-welds. The ratio of ACE to TG-43 calculated doses along the CAX (T(r)) of the 12 mm COMS plaque for the SelectSeed was on average within 3.0% of previously tabulated data, and within 2.9% of the MCNP6 simulated values. The IsoAid and SelectSeed T(r) values agreed within 0.3%. Conclusions: Initial comparisons show good agreement between ACE and MC doses for a single seed in a 12 mm COMS eye plaque; more complicated scenarios are being investigated to determine the accuracy of this calculation method.

Analysis of rainfall-induced shallow landslides in Jamne and Jaszcze stream valleys (Polish Carpathians – preliminary results

Directory of Open Access Journals (Sweden)

Zydroń Tymoteusz

2016-03-01

Full Text Available Analysis of rainfall-induced shallow landslides in Jamne and Jaszcze stream valleys (Polish Carpathians - preliminary results. Preliminary shallow landslide susceptibility mapping of the Jamne and Jaszcze stream valleys, located in the Polish Flysch Carpathians, is presented in the paper. For the purpose of mapping, there were used SINMAP and Iverson’s models integrating infiltration and slope stability calculations. The calibration of the used models parameters, obtained from limited field and laboratory tests, was performed using data from 8-9 July 1997, when as a consequence of a very intense rainfall, 94 shallow landslides were observed on meadows and arable lands. A comparison of the slope stability calculation results and the localisation of the noticed shallow landslides showed satisfactory agreement between localisation of the observed and computed unstable areas. However, it was concluded that better simulation results were obtained using Iverson’s model.

A PRELIMINARY JUPITER MODEL

International Nuclear Information System (INIS)

Hubbard, W. B.; Militzer, B.

2016-01-01

In anticipation of new observational results for Jupiter's axial moment of inertia and gravitational zonal harmonic coefficients from the forthcoming Juno orbiter, we present a number of preliminary Jupiter interior models. We combine results from ab initio computer simulations of hydrogen–helium mixtures, including immiscibility calculations, with a new nonperturbative calculation of Jupiter's zonal harmonic coefficients, to derive a self-consistent model for the planet's external gravity and moment of inertia. We assume helium rain modified the interior temperature and composition profiles. Our calculation predicts zonal harmonic values to which measurements can be compared. Although some models fit the observed (pre-Juno) second- and fourth-order zonal harmonics to within their error bars, our preferred reference model predicts a fourth-order zonal harmonic whose absolute value lies above the pre-Juno error bars. This model has a dense core of about 12 Earth masses and a hydrogen–helium-rich envelope with approximately three times solar metallicity

A PRELIMINARY JUPITER MODEL

Energy Technology Data Exchange (ETDEWEB)

Hubbard, W. B. [Lunar and Planetary Laboratory, The University of Arizona, Tucson, AZ 85721 (United States); Militzer, B. [Department of Earth and Planetary Science, Department of Astronomy, University of California, Berkeley, CA 94720 (United States)

2016-03-20

In anticipation of new observational results for Jupiter's axial moment of inertia and gravitational zonal harmonic coefficients from the forthcoming Juno orbiter, we present a number of preliminary Jupiter interior models. We combine results from ab initio computer simulations of hydrogen–helium mixtures, including immiscibility calculations, with a new nonperturbative calculation of Jupiter's zonal harmonic coefficients, to derive a self-consistent model for the planet's external gravity and moment of inertia. We assume helium rain modified the interior temperature and composition profiles. Our calculation predicts zonal harmonic values to which measurements can be compared. Although some models fit the observed (pre-Juno) second- and fourth-order zonal harmonics to within their error bars, our preferred reference model predicts a fourth-order zonal harmonic whose absolute value lies above the pre-Juno error bars. This model has a dense core of about 12 Earth masses and a hydrogen–helium-rich envelope with approximately three times solar metallicity.

Nuclear Data Processing for Reactor Physics Calculation

International Nuclear Information System (INIS)

Suwoto; Zuhair; Pandiangan, Tumpal

2003-01-01

Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1 H 1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1 H 1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV

Cadangan Full Preliminary Term Asuransi Dwiguna Dengan Hukum De Moivre

OpenAIRE

Faradilla, Sherly Mutya; ', Hasriati; Nababan, Tumpal Parulian

2015-01-01

This paper discusses premium reserve endowment life insurance for years. The reserve is calculated by the method of full preliminary term based on net annual premium, with the first net annual premium and the second net annual premium . Net annual premium is affected by amount of single premium and annuity due. De Moivre law is applied to calculate the reserve.

Calculations for very low energy scattering of positrons by molecular hydrogen

Energy Technology Data Exchange (ETDEWEB)

Cooper, J.N. [School of Mathematical Sciences, University of Nottingham, Nottingham NG7 2RD (United Kingdom)], E-mail: james.cooper@maths.nottingham.ac.uk; Armour, E.A.G. [School of Mathematical Sciences, University of Nottingham, Nottingham NG7 2RD (United Kingdom)

2008-02-15

We give a progress report on ongoing calculations of phase shifts for very low energy elastic scattering of positrons by molecular hydrogen, using the generalised Kohn variational method. Further, provisional calculations of Z{sub eff} for molecular hydrogen at low energies are presented and discussed. The preliminary nature of the work is emphasised throughout.

Defining resilience: A preliminary integrative literature review

Science.gov (United States)

Wilt, Bonnie; Long, Suzanna K.; Shoberg, Thomas G.

2016-01-01

The term “resilience” is ubiquitous in technical literature; it appears in numerous forms, such as resilience, resiliency, or resilient, and each use may have a different definition depending on the interpretation of the writer. This creates difficulties in understanding what is meant by ‘resilience’ in any given use case, especially in discussions of interdisciplinary research. To better understand this problem, this research constructs a preliminary integrative literature review to map different definitions, applications and calculation methods of resilience invoked within critical infrastructure applications. The preliminary review uses a State-of-the-Art Matrix (SAM) analysis to characterize differences in definition across disciplines and between regions. Qualifying the various usages of resilience will produce a greater precision in the literature and a deeper insight into types of data required for its evaluation, particularly with respect to critical infrastructure calculations and how such data may be analyzed. Results from this SAM analysis will create a framework of key concepts as part of the most common applications for “resilient critical infrastructure” modeling.

Simplified shielding calculation system for high-intensity proton accelerators

Energy Technology Data Exchange (ETDEWEB)

Masumura, Tomomi; Nakashima, Hiroshi; Nakane, Yoshihiro; Sasamoto, Nobuo [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

2000-06-01

A simplified shielding calculation system is developed for applying conceptual shielding design of facilities in the joint project for high-intensity proton accelerators. The system is composed of neutron transmission calculation part for bulk shielding using simplified formulas: Moyer model and Tesch's formula, and neutron skyshine calculation part using an empirical formula: Stapleton's formula. The system is made with the Microsoft Excel software for user's convenience. This report provides a manual for the system as well as calculation conditions used in the calculation such as Moyer model's parameters. In this report preliminary results based on data at December 8, 1999, are also shown as an example. (author)

Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

International Nuclear Information System (INIS)

Preece, R.J.; Putney, J.M.

1993-07-01

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

Technical note: Rapid calculation of genomic evaluations for new animals.

Science.gov (United States)

Wiggans, G R; VanRaden, P M; Cooper, T A

2015-03-01

A method was developed to calculate preliminary genomic evaluations daily or weekly before the release of official monthly evaluations by processing only newly genotyped animals using estimates of single nucleotide polymorphism effects from the previous official evaluation. To minimize computing time, reliabilities and genomic inbreeding are not calculated, and fixed weights are used to combine genomic and traditional information. Correlations of preliminary and September official monthly evaluations for animals with genotypes that became usable after the extraction of genotypes for August 2014 evaluations were >0.99 for most Holstein traits. Correlations were lower for breeds with smaller population size. Earlier access to genomic evaluations benefits producers by enabling earlier culling decisions and genotyping laboratories by making workloads more uniform across the month. Copyright © 2015 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Particle-in-Cell Calculations of the Electron Cloud in the ILC Positron Damping Ring Wigglers

International Nuclear Information System (INIS)

Celata, C.M.; Furman, M.A.; Vay, J.-L.; Grote, D.P.

2007-01-01

The self-consistent code suite WARP-POSINST is being used to study electron cloud effects in the ILC positron damping ring wiggler. WARP is a parallelized, 3D particle-in-cell code which is fully self-consistent for all species. The POSINST models for the production of photoelectrons and secondary electrons are used to calculate electron creation. Mesh refinement and a moving reference frame for the calculation will be used to reduce the computer time needed by several orders of magnitude. We present preliminary results for cloud buildup showing 3D electron effects at the nulls of the vertical wiggler field. First results from a benchmark of WARP-POSINST vs. POSINST are also discussed

Staggering towards a calculation of weak amplitudes

Energy Technology Data Exchange (ETDEWEB)

Sharpe, S.R.

1988-09-01

An explanation is given of the methods required to calculate hadronic matrix elements of the weak Hamiltonians using lattice QCD with staggered fermions. New results are presented for the 1-loop perturbative mixing of the weak interaction operators. New numerical techniques designed for staggered fermions are described. A preliminary result for the kaon B parameter is presented. 24 refs., 3 figs.

Preliminary design of RDE feedwater pump impeller

International Nuclear Information System (INIS)

Sri Sudadiyo

2018-01-01

Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

Comparison of SOLA-FLX calculations with experiments at systems, science and software

International Nuclear Information System (INIS)

Dienes, J.K.; Hirt, C.W.; Stein, L.R.

1977-03-01

Preliminary results of a comparison between hydroelastic calculations at the Los Alamos Scientific Laboratory and experiments at Systems, Science and Software are described. The axisymmetric geometry is an idealization of a pressurized water reactor at a scale of 1/25. Reasons for some of the discrepancies are described, and suggestions for improving both experiments and calculations are discussed

East Area Irradiation Test Facility: Preliminary FLUKA calculations

CERN Document Server

Lebbos, E; Calviani, M; Gatignon, L; Glaser, M; Moll, M; CERN. Geneva. ATS Department

2011-01-01

In the framework of the Radiation to Electronics (R2E) mitigation project, the testing of electronic equipment in a radiation field similar to the one occurring in the LHC tunnel and shielded areas to study its sensitivity to single even upsets (SEU) is one of the main topics. Adequate irradiation test facilities are therefore required, and one installation is under consideration in the framework of the PS East area renovation activity. FLUKA Monte Carlo calculations were performed in order to estimate the radiation field which could be obtained in a mixed field facility using the slowly extracted 24 GeV/c proton beam from the PS. The prompt ambient dose equivalent as well as the equivalent residual dose rate after operation was also studied and results of simulations are presented in this report.

Calculating computer-generated optical elements to produce arbitrary intensity distributions

International Nuclear Information System (INIS)

Findlay, S.; Nugent, K.A.; Scholten, R.E.

2000-01-01

Full text: We describe preliminary investigation into using a computer to generate optical elements (CGOEs) with phase-only variation, that will produce an arbitrary intensity distribution in a given image plane. An iterative calculation cycles between the CGOE and the image plane and modifies each according to the appropriate constraints. We extend this to the calculation of defined intensity distributions in two separated planes by modifying both phase and intensity at the CGOE

Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

International Nuclear Information System (INIS)

Olson, W.O.; Nigg, D.W.

1981-05-01

The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

Modelling lung cancer due to radon and smoking in WISMUT miners: Preliminary results

International Nuclear Information System (INIS)

Bijwaard, H.; Dekkers, F.; Van Dillen, T.

2011-01-01

A mechanistic two-stage carcinogenesis model has been applied to model lung-cancer mortality in the largest uranium-miner cohort available. Models with and without smoking action both fit the data well. As smoking information is largely missing from the cohort data, a method has been devised to project this information from a case-control study onto the cohort. Model calculations using 256 projections show that the method works well. Preliminary results show that if an explicit smoking action is absent in the model, this is compensated by the values of the baseline parameters. This indicates that in earlier studies performed without smoking information, the results obtained for the radiation parameters are still valid. More importantly, the inclusion of smoking-related parameters shows that these mainly influence the later stages of lung-cancer development. (authors)

Calculation of the Nucleon Axial Form Factor Using Staggered Lattice QCD

Energy Technology Data Exchange (ETDEWEB)

Meyer, Aaron S. [Fermilab; Hill, Richard J. [Perimeter Inst. Theor. Phys.; Kronfeld, Andreas S. [Fermilab; Li, Ruizi [Indiana U.; Simone, James N. [Fermilab

2016-10-14

The nucleon axial form factor is a dominant contribution to errors in neutrino oscillation studies. Lattice QCD calculations can help control theory errors by providing first-principles information on nucleon form factors. In these proceedings, we present preliminary results on a blinded calculation of $g_A$ and the axial form factor using HISQ staggered baryons with 2+1+1 flavors of sea quarks. Calculations are done using physical light quark masses and are absolutely normalized. We discuss fitting form factor data with the model-independent $z$ expansion parametrization.

Preliminary GRS Measurement of Chlorine Distribution on Surface of Mars

Science.gov (United States)

Keller, J. M.; Boynton, W. V.; Taylor, G. J.; Hamara, D.; Janes, D. M.; Kerry, K.

2003-12-01

Ongoing measurements with the Gamma Ray Spectrometer (GRS) aboard Mars Odyssey provide preliminary detection of chlorine at the surface of Mars. Summing all data since boom deployment and using a forward calculation model, we estimate values for chlorine concentration at 5° resolution. Rebinning this data and smoothing with a 15-degree-radius boxcar filter reveal regions of noticeable chlorine enrichment at scales larger than the original 5° resolution and allow for preliminary comparison with previous Mars datasets. Analyzing chlorine concentrations within 30 degrees of the equator, we find a negative correlation with thermal inertia (R2=0.55) and positive correlation with albedo (R2=0.52), indicating that chlorine is associated with fine, non-rock surface materials. Although possibly a smoothing artifact, the spatial correlation is more noticeable in the region covering Tharsis and Amazonis than around Arabia and Elysium. Additionally, a noticeable region of chlorine enrichment appears west of Tharsis Montes ( ˜0 to 20N, ˜110 to 150W) and chlorine concentration is estimated to vary in the equatorial region by over a factor of two. A simplified two-component model involving chlorine-poor rocks and a homogenous chlorine-rich fine material requires rock abundance to vary from zero to over 50%, a result inconsistent with previous measurements and models. In addition to variations in rock composition and distribution, substantial variations in chlorine content of various types of fine materials including dust, sand, and duricrust appear important in explaining this preliminary observation. Surprisingly, visual comparison of surface units mapped by Christensen and Moore (1992) does not show enrichment in chlorine associated with regions of indurated surfaces, where cementation has been proposed. Rather, Tharsis, a region of active deposition with proposed mantling of 0.1 to 2 meters of recent dust (Christensen 1986), shows the greatest chlorine signal. In light of

A preliminary model for estimating the first wall lifetime of a fusion reactor

International Nuclear Information System (INIS)

Daenner, W.

1975-02-01

The estimation of the first wall lifetime is a necessary basis for predicting the availability of a fusion power plant. In order to do this, an analytical model was prepared and programmed for the computer which calculates the temperature and stress load of the first wall from the principal design parameters and quotes them against the relevant material properties. Neither the analytical model nor the information about the material performance is yet complete so that the answers obtained from the program are very preliminary. This situation is underlined by the results of sample calculations performed for the CTRD blanket module cell. The results obtained for vanadium and vanadium alloys show a strong dependence of the lifetime on the irradiation creep and the ductility of these materials. Completion of this model is envisaged as soon as the missing information becomes available. (orig.) [de

Technetium removal: preliminary flowsheet options

International Nuclear Information System (INIS)

Eager, K.M.

1995-01-01

This document presents the results of a preliminary investigation into options for preliminary flowsheets for 99Tc removal from Hanford Site tank waste. A model is created to show the path of 99Tc through pretreatment to disposal. The Tank Waste Remediation (TWRS) flowsheet (Orme 1995) is used as a baseline. Ranges of important inputs to the model are developed, such as 99Tc inventory in the tanks and important splits through the TWRS flowsheet. Several technetium removal options are discussed along with sensitivities of the removal schemes to important model parameters

Artificial neural networks, a new alternative to Monte Carlo calculations for radiotherapy

International Nuclear Information System (INIS)

Martin, E.; Gschwind, R.; Henriet, J.; Sauget, M.; Makovicka, L.

2010-01-01

In order to reduce the computing time needed by Monte Carlo codes in the field of irradiation physics, notably in dosimetry, the authors report the use of artificial neural networks in combination with preliminary Monte Carlo calculations. During the learning phase, Monte Carlo calculations are performed in homogeneous media to allow the building up of the neural network. Then, dosimetric calculations (in heterogeneous media, unknown by the network) can be performed by the so-learned network. Results with an equivalent precision can be obtained within less than one minute on a simple PC whereas several days are needed with a Monte Carlo calculation

Calculation of 235U(n,n') cross sections for ENDF/B-VI

International Nuclear Information System (INIS)

Young, P.G.; Arthur, E.D.

1988-01-01

Cross sections for neutron-induced reactions on 235 U between 0.01 and 20 MeV have been calculated in a preliminary analysis for the ENDF/B-VI evaluation with particular emphasis on neutron inelastic scattering. A deformed optical model potential that fits total, elastic, inelastic, and low-energy average resonance data is used to calculate direct (n,n') cross sections and transmission coefficients for a Hauser-Feshbach statistical theory analysis using a multiple fission barrier representation. Direct cross sections for higher-lying vibrational states are provided from DWBA calculations, normalized using B(E/ital l/) values determined from (d,d') and Coulomb excitation data. Initial fission barrier parameters and transition state density enhancements appropriate to the compound systems involved were obtained from previous analyses, especially fits to charged-particle fission probability data. Further modifications to fit 235 U(n,f) data were small, and the final fission parameters are generally consistent with published values. The results from this preliminary analysis are compared with the ENDF/B-V evaluation as well as with experimental data. 26 refs., 5 figs., 3 tabs

Preliminary calculations of release rates from spent fuel in a tuff repository

International Nuclear Information System (INIS)

Apted, M.J.; O'Connell, W.J.; Lee, K.H.; MacIntyre, A.T.; Ueng, T.S.; Pigford, T.H.; Lee, W.W.L.

1991-01-01

Time-dependent release rates of Tc-99, I-129, Cs-135, and Np-237 have been calculated for wet-drip and moist-continuous release modes from the engineered barrier system of a potential nuclear waste repository in unsaturated tuff, representative of a possible repository at Yucca Mountain in southern Nevada. We describe the modes of water contact and of release of dissolved radionuclides to the surrounding intact rock, and the corresponding calculational models. We list the parameter values adopted, and then present numerical results, conclusions, and recommendations. 21 refs., 5 figs., 2 tabs

Preliminary deformation model for National Seismic Hazard map of Indonesia

Energy Technology Data Exchange (ETDEWEB)

Meilano, Irwan; Gunawan, Endra; Sarsito, Dina; Prijatna, Kosasih; Abidin, Hasanuddin Z. [Geodesy Research Division, Faculty of Earth Science and Technology, Institute of Technology Bandung (Indonesia); Susilo,; Efendi, Joni [Agency for Geospatial Information (BIG) (Indonesia)

2015-04-24

Preliminary deformation model for the Indonesia’s National Seismic Hazard (NSH) map is constructed as the block rotation and strain accumulation function at the elastic half-space. Deformation due to rigid body motion is estimated by rotating six tectonic blocks in Indonesia. The interseismic deformation due to subduction is estimated by assuming coupling on subduction interface while deformation at active fault is calculated by assuming each of the fault‘s segment slips beneath a locking depth or in combination with creeping in a shallower part. This research shows that rigid body motion dominates the deformation pattern with magnitude more than 15 mm/year, except in the narrow area near subduction zones and active faults where significant deformation reach to 25 mm/year.

Nuclide release calculation in the near-field of a reference HLW repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

2004-01-01

The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997 in order to develop a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as waste and generic site characteristics in Korea was roughly envisaged in 2003 focusing on the near-field components of the repository system. According to above basic repository concept, which is similar to that of Swedish KBS-3 repository, the spent fuel is first encapsulated in corrosion resistant canisters, even though the material has not yet been determined, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels constructed at a depth of about 500 m in a stable plutonic rock body. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near-field components of the repository, even though enough information has not been available that much yet, but also to show a methodology by which a generic safety assessment could be performed for further development of Korea reference repository concept, nuclide release calculation study strongly seems to be necessary

Meta-GWAS Accuracy and Power (MetaGAP Calculator Shows that Hiding Heritability Is Partially Due to Imperfect Genetic Correlations across Studies.

Directory of Open Access Journals (Sweden)

Ronald de Vlaming

2017-01-01

Full Text Available Large-scale genome-wide association results are typically obtained from a fixed-effects meta-analysis of GWAS summary statistics from multiple studies spanning different regions and/or time periods. This approach averages the estimated effects of genetic variants across studies. In case genetic effects are heterogeneous across studies, the statistical power of a GWAS and the predictive accuracy of polygenic scores are attenuated, contributing to the so-called 'missing heritability'. Here, we describe the online Meta-GWAS Accuracy and Power (MetaGAP calculator (available at www.devlaming.eu which quantifies this attenuation based on a novel multi-study framework. By means of simulation studies, we show that under a wide range of genetic architectures, the statistical power and predictive accuracy provided by this calculator are accurate. We compare the predictions from the MetaGAP calculator with actual results obtained in the GWAS literature. Specifically, we use genomic-relatedness-matrix restricted maximum likelihood to estimate the SNP heritability and cross-study genetic correlation of height, BMI, years of education, and self-rated health in three large samples. These estimates are used as input parameters for the MetaGAP calculator. Results from the calculator suggest that cross-study heterogeneity has led to attenuation of statistical power and predictive accuracy in recent large-scale GWAS efforts on these traits (e.g., for years of education, we estimate a relative loss of 51-62% in the number of genome-wide significant loci and a relative loss in polygenic score R2 of 36-38%. Hence, cross-study heterogeneity contributes to the missing heritability.

Preliminary neutron shielding calculations of the electronics in the EAST BES systems focusing on neutron induced displacement damage

Energy Technology Data Exchange (ETDEWEB)

Náfrádi, Gábor, E-mail: nafradi@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Kovácsik, Ákos, E-mail: kovacsik.akos@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Németh, József, E-mail: nemeth.jozsef@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary); Pór, Gábor, E-mail: por@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Zoletnik, Sándor, E-mail: zoletnik.sandor@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary)

2016-11-15

Monte Carlo N-Particle (MCNP) calculations were carried out to compare neutron shielding capabilities of three frequently used neutron shielding materials: polyethylene without neutron absorbers, polyethylene with boron absorbers and polyethylene with lithium absorbers, according to Non Ionizing Energy Loss (NIEL). The results of 1D shielding calculations showed that simple neutron moderating materials can provide sufficient and cheap shielding against 2.45 MeV and 14.1 MeV fusion neutrons, in terms of 1 MeV neutron equivalent flux, in silicon targets, which is the most commonly used material of electronic components. Based on these results a new shielding concept is proposed which can be taken into consideration where the reduction of displacement damage is the main goal and the free space available for shielding is limited. Based on this shielding concept detailed 3D calculations were carried out to describe the properties of the neutron shielding of the Beam Emission Spectroscopy (BES) system installed at the EAST tokamak.

Ionizing radiation calculations and comparisons with LDEF data

Science.gov (United States)

Armstrong, T. W.; Colborn, B. L.; Watts, J. W., Jr.

1992-01-01

In conjunction with the analysis of LDEF ionizing radiation dosimetry data, a calculational program is in progress to aid in data interpretation and to assess the accuracy of current radiation models for future mission applications. To estimate the ionizing radiation environment at the LDEF dosimeter locations, scoping calculations for a simplified (one dimensional) LDEF mass model were made of the primary and secondary radiations produced as a function of shielding thickness due to trapped proton, galactic proton, and atmospheric (neutron and proton cosmic ray albedo) exposures. Preliminary comparisons of predictions with LDEF induced radioactivity and dose measurements were made to test a recently developed model of trapped proton anisotropy.

Project W-320, 241-C-106 sluicing: Civil/structural calculations. Volume 6

Energy Technology Data Exchange (ETDEWEB)

Bailey, J.W.

1998-07-24

This supporting document has been prepared to make the FDNW calculations for Project W-320 readily retrievable. The purpose of this calculation is to conservatively estimate the weight of equipment and structures being added over Tank 241-C-106 as a result of Project W-320 and combine these weights with the estimated weights of existing structures and equipment as calculated in Attachment 1. The combined weights will be compared to the allowable live load limit to provide a preliminary assessment of loading conditions above Tank 241-C-106.

Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

2017-07-01

In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

Approximate Method of Calculating Forces on Rudder During Ship Sailing on a Shipping Route

Directory of Open Access Journals (Sweden)

K. Zelazny

2014-09-01

Full Text Available Service speed of a ship in real weather conditions is a basic design parameter. Forecasting of this speed at preliminary design stage is made difficult by the lack of simple but at the same accurate models of forces acting upon a ship sailing on a preset shipping route. The article presents a model for calculating forces and moment on plane rudder, useful for forecasting of ship service speed at preliminary stages of ship design.

Core design calculations of IRIS reactor using modified CORD-2 code package

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

2002-01-01

Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

Preliminary Design of a LSA Aircraft Using Wind Tunnel Tests

Directory of Open Access Journals (Sweden)

Norbert ANGI

2015-12-01

Full Text Available This paper presents preliminary results concerning the design and aerodynamic calculations of a light sport aircraft (LSA. These were performed for a new lightweight, low cost, low fuel consumption and long-range aircraft. The design process was based on specific software tools as Advanced Aircraft Analysis (AAA, XFlr 5 aerodynamic and dynamic stability analysis, and Catia design, according to CS-LSA requirements. The calculations were accomplished by a series of tests performed in the wind tunnel in order to assess experimentally the aerodynamic characteristics of the airplane.

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)

Preliminary neutronic design of spock reactor: A nuclear system for space power generation

International Nuclear Information System (INIS)

Burgio, N.; Santagata, A.; Cumo, M.; Fasano, A.; Frullini, M.

2007-01-01

the feasibility of the standard nuclear data sets for the design of space nuclear devices. Finally, FLUENT 6.2.16 preliminary calculations show for a hot pin fuel temperature of 1700 K, 325 kW t h were transferred to the coolant (Γ= 1.8 Kg/s, ΔT = 150 K) with the possibility to be converted in 30 kW e l by using an advanced thermoelectric converter system. Reference: [1] M. Cumo, M. Frullini, A. Gandini, A. Naviglio, L. Sorabella 'MAUS - 1.5 Nuclear Reactor for Space Electric Power', ICENES 2005- Bruxelles August 2005

Error estimates for ice discharge calculated using the flux gate approach

Science.gov (United States)

Navarro, F. J.; Sánchez Gámez, P.

2017-12-01

Ice discharge to the ocean is usually estimated using the flux gate approach, in which ice flux is calculated through predefined flux gates close to the marine glacier front. However, published results usually lack a proper error estimate. In the flux calculation, both errors in cross-sectional area and errors in velocity are relevant. While for estimating the errors in velocity there are well-established procedures, the calculation of the error in the cross-sectional area requires the availability of ground penetrating radar (GPR) profiles transverse to the ice-flow direction. In this contribution, we use IceBridge operation GPR profiles collected in Ellesmere and Devon Islands, Nunavut, Canada, to compare the cross-sectional areas estimated using various approaches with the cross-sections estimated from GPR ice-thickness data. These error estimates are combined with those for ice-velocities calculated from Sentinel-1 SAR data, to get the error in ice discharge. Our preliminary results suggest, regarding area, that the parabolic cross-section approaches perform better than the quartic ones, which tend to overestimate the cross-sectional area for flight lines close to the central flowline. Furthermore, the results show that regional ice-discharge estimates made using parabolic approaches provide reasonable results, but estimates for individual glaciers can have large errors, up to 20% in cross-sectional area.

Preliminary parameter assessments of a spiral FFAG accelerator for proton therapy

International Nuclear Information System (INIS)

Smirnov, V.L.; Azaryan, N.S.; Vorozhtsov, S.B.

2013-01-01

Fixed-Field Alternating-Gradient (FFAG) accelerator was invented in the 1950-60s but never progressed beyond the model stage. Starting from 2000, new interest in this type of accelerator arose. Given advantages of the FFAG over the synchrotron, cyclotron and linac, there are many possible applications of the accelerator. Among them, we are mostly interested in acceleration of protons and light ions for hadron therapy. In this connection a preliminary set of parameters of the facility was estimated and, in particular, the magnetic sector shape and corresponding dynamical properties of the magnetic field of the accelerator were calculated. In addition, preliminary considerations about the RF system design are given.

Preliminary study to improve the performance of SCWR-M during loss-of-flow accident

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Sun, C.; Wang, Z.D.; Chai, X.; Xiong, J.B.; Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

2016-10-15

Highlights: • Validation of the ATHLET-SC code to the safety analysis for SCWR. • Loss of flow accident analysis for SCWR-M is performed. • The passive design parameter is optimized. • The optimized SCWR-M design shows a better safety performance. - Abstract: The SCWR-M is one of the conceptual core designs with mixed neutron spectrum (fast and thermal), which is developed at Shanghai Jiao Tong University. Some preliminary calculations of this new conceptual SCWR indicate the SCWR-M system gets better safety characteristics compared to other single spectrum supercritical water cooled reactors. Loss of flow accident (LOFA) is of particular importance among the abnormal events and accidents for SCWR-M. In order to perform the preliminary study to improve the current SCWR-M safety design, this paper presents the validation results of the ATHLET-SC code and optimization work for safety system design parameters of the ICS, ACC, GDCS based on LOFA analysis. The better performance of the optimized design parameters are demonstrated by comparison with the previous design.

How to calculate linear absorption spectra with lifetime broadening using fewest switches surface hopping trajectories: A simple generalization of ground-state Kubo theory

International Nuclear Information System (INIS)

Petit, Andrew S.; Subotnik, Joseph E.

2014-01-01

In this paper, we develop a surface hopping approach for calculating linear absorption spectra using ensembles of classical trajectories propagated on both the ground and excited potential energy surfaces. We demonstrate that our method allows the dipole-dipole correlation function to be determined exactly for the model problem of two shifted, uncoupled harmonic potentials with the same harmonic frequency. For systems where nonadiabatic dynamics and electronic relaxation are present, preliminary results show that our method produces spectra in better agreement with the results of exact quantum dynamics calculations than spectra obtained using the standard ground-state Kubo formalism. As such, our proposed surface hopping approach should find immediate use for modeling condensed phase spectra, especially for expensive calculations using ab initio potential energy surfaces

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)

CALCULATION ALGORITHM TRUSS UNDER CRANE BEAMS

Directory of Open Access Journals (Sweden)

N. K. Akaev1

2016-01-01

Full Text Available Aim.The task of reducing the deflection and increase the rigidity of single-span beams are made. In the article the calculation algorithm for truss crane girders is determined.Methods. To identify the internal effort required for the selection of cross section elements the design uses the Green's function.Results. It was found that the simplest truss system reduces deflection and increases the strength of design. The upper crossbar is subjected not only to bending and shear and compression work due to tightening tension. Preliminary determination of the geometrical characteristics of the crane farms elements are offered to make a comparison with previous similar configuration of his farms, using a simple approximate calculation methods.Conclusion.The method of sequential movements (incrementally the two bridge cranes along the length of the upper crossbar truss beams is suggested. We give the corresponding formulas and conditions of safety.

Application of backtracking algorithm to depletion calculations

International Nuclear Information System (INIS)

Wu Mingyu; Wang Shixi; Yang Yong; Zhang Qiang; Yang Jiayin

2013-01-01

Based on the theory of linear chain method for analytical depletion calculations, the burnup matrix is decoupled by the divide and conquer strategy and the linear chain with Markov characteristic is formed. The density, activity and decay heat of every nuclide in the chain then can be calculated by analytical solutions. Every possible reaction path of the nuclide must be considered during the linear chain establishment process. To confirm the calculation precision and efficiency, the algorithm which can cover all the reaction paths and search the paths automatically according to the problem description and precision restrictions should be found. Through analysis and comparison of several kinds of searching algorithms, the backtracking algorithm was selected to establish and calculate the linear chains in searching process using depth first search (DFS) method, forming an algorithm which can solve the depletion problem adaptively and with high fidelity. The complexity of the solution space and time was analyzed by taking into account depletion process and the characteristics of the backtracking algorithm. The newly developed depletion program was coupled with Monte Carlo program MCMG-Ⅱ to calculate the benchmark burnup problem of the first core of China Experimental Fast Reactor (CEFR) and the preliminary verification and validation of the program were performed. (authors)

Preliminary Calculation of the Indicators of Sustainable Development for National Radioactive Waste Management Programs

International Nuclear Information System (INIS)

Cheong, Jae Hak; Park, Won Jae

2003-01-01

As a follow up to the Agenda 21's policy statement for safe management of radioactive waste adopted at Rio Conference held in 1992, the UN invited the IAEA to develop and implement indicators of sustainable development for the management of radioactive waste. The IAEA finalized the indicators in 2002, and is planning to calculate the member states' values of indicators in connection with operation of its Net-Enabled Waste Management Database system. In this paper, the basis for introducing the indicators into the radioactive waste management was analyzed, and calculation methodology and standard assessment procedure were simply depicted. In addition, a series of innate limitations in calculation and comparison of the indicators was analyzed. According to the proposed standard procedure, the indicators for a few major countries including Korea were calculated and compared, by use of each country's radioactive waste management framework and its practices. In addition, a series of measures increasing the values of the indicators was derived so as to enhance the sustainability of domestic radioactive waste management program.

Evaluation of calculational and material models for concrete containment structures

International Nuclear Information System (INIS)

Dunham, R.S.; Rashid, Y.R.; Yuan, K.A.

1984-01-01

A computer code utilizing an appropriate finite element, material and constitutive model has been under development as a part of a comprehensive effort by the Electric Power Research Institute (EPRI) to develop and validate a realistic methodology for the ultimate load analysis of concrete containment structures. A preliminary evaluation of the reinforced and prestressed concrete modeling capabilities recently implemented in the ABAQUS-EPGEN code has been completed. This effort focuses on using a state-of-the-art calculational model to predict the behavior of large-scale reinforced concrete slabs tested under uniaxial and biaxial tension to simulate the wall of a typical concrete containment structure under internal pressure. This paper gives comparisons between calculations and experimental measurements for a uniaxially-loaded specimen. The calculated strains compare well with the measured strains in the reinforcing steel; however, the calculations gave diffused cracking patterns that do not agree with the discrete cracking observed in the experiments. Recommendations for improvement of the calculational models are given. (orig.)

Vivitron 1995, transient voltage simulation, high voltage insulator tests, electric field calculation

International Nuclear Information System (INIS)

Frick, G.; Osswald, F.; Heusch, B.

1996-01-01

Preliminary investigations showed clearly that, because of the discrete electrode structure of the Vivitron, important overvoltage leading to insulator damage can appear in case of a spark. The first high voltage tests showed damage connected with such events. This fact leads to a severe voltage limitation. This work describes, at first, studies made to understand the effects of transients and the associated over-voltage appearing in the Vivitron. Then we present the high voltage tests made with full size Vivitron components using the CN 6 MV machine as a pilot machine. Extensive field calculations were made. These involve simulations of static stresses and transient overvoltages, on insulating boards and electrodes. This work gave us the solutions for arrangements and modifications in the machine. After application, the Vivitron runs now without any sparks and damage at 20 MV. In the same manner, we tested column insulators of a new design and so we will find out how to get to higher voltages. Electric field calculation around the tie bars connecting the discrete electrodes together showed field enhancements when the voltages applied on the discrete electrodes are not equally distributed. This fact is one of the sources of discharges and voltage limitations. A scenario of a spark event is described and indications are given how to proceed towards higher voltages, in the 30 MV range. (orig.)

Preliminary data evaluation for thermal insulation characterization testing

International Nuclear Information System (INIS)

DeClue, J.F.; Moses, S.D.; Tollefson, D.A.

1991-01-01

The purpose of Thermal Insulation Characterization Testing is to provide physical data to support certain assumptions and calculational techniques used in the criticality safety calculations in Section 6 of the Safety Analysis Reports for Packaging (SARPs) for drum-type packaging for Department of Energy's (DOE) Oak Ridge Y-12 Plant, managed by Martin Marietta Energy Systems, Inc. Results of preliminary data evaluation regarding the fire-test condition reveal that realistic weight loss consideration and residual material characterization in developing calculational models for the hypothetical accident condition is necessary in order to prevent placement of unduly conservative restrictions on shipping requirements as a result of overly conservative modeling. This is particularly important for fast systems. Determination of the geometric arrangement of residual material is of secondary importance. Both the methodology used to determine the minimum thermal insulation mass remaining after the fire test and the treatment of the thermal insulation in the criticality safety calculational models requires additional evaluation. Specific testing to be conducted will provide experimental data with which to validate the mass estimates and calculational modeling techniques for extrapolation to generic drum-type containers

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

Radiolabelling of RC-160: preliminary results

International Nuclear Information System (INIS)

Verdera, E.S.; Balter Binsky, H.S.; Robles, A.M.; Rodriguez, G.; Souto, B.; Laiz, J.; Oliver, P.; Leon, E.

1998-01-01

Vapreotide (RC-160) was labelled with 125 I using Chloramine-T and Iodogen methods and with 99m Tc by a direct method with sodium ditionite as reducing agent in the presence of ascorbic acid. Several methods of purification and quality control were evaluated. Yields of the reactions and of purification steps were calculated. The results obtained for the radioiodination reactions showed higher yields when limiting Chloramine-T method was used. Labelling of RC-160 with 99m Tc indicated better yields when high radioactivity concentration of the radionuclide was used. Stability of the products obtained was assessed at different post-labelling times by selected quality control methods: Sep-Pak cartridge as purification method and chromatography by RP-HPLC and ITLC-SG using saline solution as solvent. It was demonstrated that I-125-RC-160 and Tc-99m-RC-160 were stable during five weeks (at -20 deg. C) and 6 hours (at room temperature) respectively. Preliminary biodistribution of Tc-99m-RC-160 in normal rats and mice were done showing different biological behaviour compared with control animals injected with pertechnetate. In conclusion, RC-160 was successfully labelled with both radionuclides, with radiochemical purity higher than 95%. These results encourage further research work in animal models as well as to investigate the biochemical behaviour of radiolabelled peptide. (author)

Adjoint electron Monte Carlo calculations

International Nuclear Information System (INIS)

Jordan, T.M.

1986-01-01

Adjoint Monte Carlo is the most efficient method for accurate analysis of space systems exposed to natural and artificially enhanced electron environments. Recent adjoint calculations for isotropic electron environments include: comparative data for experimental measurements on electronics boxes; benchmark problem solutions for comparing total dose prediction methodologies; preliminary assessment of sectoring methods used during space system design; and total dose predictions on an electronics package. Adjoint Monte Carlo, forward Monte Carlo, and experiment are in excellent agreement for electron sources that simulate space environments. For electron space environments, adjoint Monte Carlo is clearly superior to forward Monte Carlo, requiring one to two orders of magnitude less computer time for relatively simple geometries. The solid-angle sectoring approximations used for routine design calculations can err by more than a factor of 2 on dose in simple shield geometries. For critical space systems exposed to severe electron environments, these potential sectoring errors demand the establishment of large design margins and/or verification of shield design by adjoint Monte Carlo/experiment

A Preliminary Fire PSA on PGSFR

International Nuclear Information System (INIS)

Kim, Kilyoo; Han, Sanghoon; Lee, KwiLim

2017-01-01

A Prototype Generation IV Sodium Fast Reactor (PGSFR) is under design with defense in depth concept with active, passive, and inherent safety features to acquire a design approval for PGSFR from Korean regulatory authority by around 2017. A preliminary fire PSA on PGSFR is done in 2016 and a final fire PSA of PGSFR will be done in 2017. The characteristics of the preliminary fire PSA on PGSFR are described in this paper. Since PGSFR is very safe reactor, it is not bad approach to use a conservative assumption in the preliminary PSA. In addition, several drawings including cable routing are not yet issued, a conservative calculation for CDF is performed. As shown in Table 2, the CDF caused by the fire in the control room takes 89% portion of total CDF. Thus, a detailed fire modeling for control room is necessary for the final fire PSA on PGSFR. Also, the increased ignition frequency due to sodium leak would be derived by considering the sodium piping complexity in the final fire PSA on PGSFR. The 4th column of Table 2 is derived the 3rd column by multiplying the factor (592/1177). The 5th column is the ignition frequency caused by the sodium leak. The 6th column is derived by summing the 4th column and the 5th column. The 7th column is the CDF portion of each fire area. The control room (fire area F-A404A) is the most important area since the control room fire takes 89% portion of total CDF.

Calculation of magnetic field and electromagnetic forces in MHD superconducting magnets

International Nuclear Information System (INIS)

Martinelli, G.; Morini, A.; Moisio, M.F.

1992-01-01

The realization of a superconducting prototype magnet for MHD energy conversion is under development in Italy. Electromechanical industries and University research groups are involved in the project. The paper deals with analytical methods developed at the Department of Electrical Engineering of Padova University for calculating magnetic field and electromagnetic forces in MHD superconducting magnets and utilized in the preliminary design of the prototype

Radioactive waste shredding: Preliminary evaluation

International Nuclear Information System (INIS)

Soelberg, N.R.; Reimann, G.A.

1994-07-01

The critical constraints for sizing solid radioactive and mixed wastes for subsequent thermal treatment were identified via a literature review and a survey of shredding equipment vendors. The types and amounts of DOE radioactive wastes that will require treatment to reduce the waste volume, destroy hazardous organics, or immobilize radionuclides and/or hazardous metals were considered. The preliminary steps of waste receipt, inspection, and separation were included because many potential waste treatment technologies have limits on feedstream chemical content, physical composition, and particle size. Most treatment processes and shredding operations require at least some degree of feed material characterization. Preliminary cost estimates show that pretreatment costs per unit of waste can be high and can vary significantly, depending on the processing rate and desired output particle size

Calculation of the beam injector steering system using Helmholtz coils

International Nuclear Information System (INIS)

Passaro, A.; Sircilli Neto, F.; Migliano, A.C.C.

1991-03-01

In this work, a preliminary evaluation of the beam injector steering system of the IEAv electron linac is presented. From the existing injector configuration and with the assumptions of monoenergetic beam (100 keV) and uniform magnetic field, two pairs of Helmholtz coils were calculated for the steering system. Excitations of 105 A.turn and 37 A.turn were determined for the first and second coils, respectively. (author)

A dielectric matrix calculation of the surface-plasmon energy for the silicon (100) surface

International Nuclear Information System (INIS)

Forsyth, A.J.; Smith, A.E.; Josefsson, T.W.

1996-01-01

Full text: As an extension of previous work, we present preliminary calculations for the dielectric properties of the silicon (100) surface. In particular, the |q|→0 and |q|=2π/a(1,0,0) surface loss function, and corresponding surface plasmon energies have been calculated within a simple model for the silicon surface. The results have been obtained from the Adler and Wiser dielectric matrix (DM). The bandstructure used for the calculation was based on the highly successful empirical pseudopotential method of Cohen and Chelikovsky. We have used a 59 plane wave basis for the bandstructure, and have chosen a DM size of 59 x 59. Results are compared and contrasted with volume plasmon calculations, free electron calculations and experiment

Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

Microscopic calculation of four-nucleon scattering observables in dd → dd and dd → p3H

International Nuclear Information System (INIS)

Fonseca, A.C.

1998-01-01

The four-body equations of Alt, Grassberger and Sandhas are solved for a system of four nucleons, using realistic NN interactions. The results of the calculations are compared with data for the reactions and dd → dd and dd → p 3 H. Preliminary calculations indicate that the nucleon-nucleon p-waves have a strong effect on 4N observables. (orig.)

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)

The 'equivalent plutonium' concept and its application to synergetic fuel cycles calculation

International Nuclear Information System (INIS)

Perez Tumini, L.L.; Sbaffoni, M.M.; Abbate, M.J.

1995-01-01

The advanced fuel cycles are seen as very interesting alternatives to improve the utilization of uranium resources in the middle term. Among them, the synergetic cycles between different type of reactors, particularly PWR and CANDU are seen as very promising. In the frame of the Argentinean-Brazilian cooperation agreement, a neutronic and economical study was done on a Tandem cycle between the Brazilian Pressurized Water Reactor Angra-I, and the Argentinean CANDU reactor Embalse. The first calculations showed very interesting results regarding the obtainable savings in natural resources, the cost of the fuel cycle, and the lower quantity of wastes to be disposed. To perform the initial calculations, two methods were mainly used: standard calculation codes, which use discrete ordinates or collision probabilities method to solve the neutronics of the cell, or an algorithm that from now on we will call EQUIVALENT PLUTONIUM. The present work describes the concept in which the algorithm is based, the obtention of the coefficients needed for its determination, and, as an example, the results obtained applying the algorithm to two particular cases of Tandem cycles: CANDU MOX fuel fabricated from PWR fuel, diluted with natural uranium, and with depleted uranium. The obtained results are compared with calculations performed with WIMS code. It was verified that the methodology which makes use of the concepts of equivalent plutonium simplifies a lot burn-up and blending radio calculations for preliminary fuel cycle analysis, giving results with very good approximation (approximately 5%) and in a very simple way. (author)

Eurisol-DS Multi MW Target Preliminary Study of the Windlowless Transverse Film (WTF) Liquid Metal Proton-to Neutron Converter

CERN Document Server

Kadi, Y; Rocca, R; Samec, K

2008-01-01

This technical note summarises the design calculations performed within Task#2 of the European Isotope Separation On-Line Radioactive Ion Beam Facility Design Study (EURISOL-DS) for the WTF (Windowless Transverse Film) mercur converter. A preliminary study was carried out in order to determine the heat deposition within the mercury and estimate the mercury velocity needed in the film. The geometry used is based on previous analysis simulated using the Monte Carlo code FLUKA. The results of these calculations show the baseline parameters, which will be used for the detailed design. Particularly, with a 1 GeV proton beam with a $\\sigma$ ~2 mm Gaussian distribution on a 4x30x40cm long target and with a 5m/s velocity at the peak power density region seems a suitable solution.

Multi-compartment iodine calculations with FIPLOC/IMPAIR

Energy Technology Data Exchange (ETDEWEB)

Ewig, F; Allelein, H J [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Schwarz, S; Weber, G [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

1996-12-01

The multi-compartment containment code FIPLOC for the simulation of severe accidents in LWR plants was extended by the integration of the iodine model IMPAIR-3. The iodine model was changed for arbitrary compartment configurations and tightly coupled to the thermal hydraulic part. A main progress with the coupled version FIPLOC-3.0 is the sophisticated modelling of the aerosol iodine behaviour. In a PWR accident the mass of iodine is mainly released in form of CsI aerosol from the primary circuit. In IMPAIR-3 the aerosol behaviour of the species CsI, AgI and IO{sub 3}{sup -} is modelled in a very simplified way causing large uncertainties in the calculated distributions. The behaviour of these three aerosol species is treated by the aerosol model MAEROS/MGA. Agglomeration, particle growth by condensation and all deposition processes are calculated. The solubility effect for the hygroscopic species CsI and IO{sub 3}{sup -} are comprehended. Furthermore the impact of the iodine decay heat on the thermal hydraulic behaviour is considered. In order to test the code development a preliminary FIPLOC-3.0 calculation was done simulating a German PWR containment for the core melt scenario ND* according to the German risk study phase B. IN the calculation a contact of the core melt with the sump water was assumed and the containment vent line was opened after 70 hours. The result show that the different iodine species are distributed inhomogeneously within the containment. The CsI-aerosol concentrations differ by two orders of magnitude and the I{sub 2}-concentration even by three orders of magnitude. Most of the iodine is assumed to be released as CsI aerosol out of the primary circuit. Since it fastly deposits its contribution to the release into the environment is minor. CsI is however dissolved in the sump, where mainly the gaseous I{sub 2} is created which can react in the containment atmosphere to IO{sub 3}{sup -}. (author) 11 figs., 3 tabs., 12 refs.

Multi-compartment iodine calculations with FIPLOC/IMPAIR

International Nuclear Information System (INIS)

Ewig, F.; Allelein, H.J.; Schwarz, S.; Weber, G.

1996-01-01

The multi-compartment containment code FIPLOC for the simulation of severe accidents in LWR plants was extended by the integration of the iodine model IMPAIR-3. The iodine model which originally was only drafted for chains of compartments was changed for arbitrary compartment configurations and tightly coupled to the thermal hydraulic part. A main progress with the coupled version FIPLOC-3.0 is the sophisticated modelling of the aerosol iodine behaviour. In a PWR accident the mass of iodine is mainly released in form of CsI aerosol from the primary circuit. In IMPAIR-3 the aerosol behaviour of the species CsI, AgI and IO 3 - is modelled in a very simplified way causing large uncertainties in the calculated distributions. The behaviour of these three aerosol species is treated by the aerosol model MAEROS/MGA. Agglomeration, particle growth by condensation and all deposition processes are calculated. The solubility effect for the hygroscopic species CsI and IO 3 - are comprehended. Furthermore the impact of the iodine decay heat on the thermal hydraulic behaviour is considered. In order to test the code development a preliminary FIPLOC-3.0 calculation was done simulating a German PWR containment for the core melt scenario ND* according to the German risk study phase B. IN the calculation a contact of the core melt with the sump water was assumed and the containment vent line was opened after 70 hours. The result show that the different iodine species are distributed inhomogeneously within the containment. The CsI-aerosol concentrations differ by two orders of magnitude and the I 2 -concentration even by three orders of magnitude. Most of the iodine is assumed to be released as CsI aerosol out of the primary circuit. Since it fastly deposits its contribution to the release into the environment is minor. CsI is however dissolved in the sump, where mainly the gaseous I 2 is created which can react in the containment atmosphere to IO 3 - . (author) 11 figs., 3 tabs., 12

Methods of statistical calculation of fast reactor core with account of influence of fuel assembly form change in process of campaign and other factors

International Nuclear Information System (INIS)

Sorokin, G.A.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

2000-01-01

The method of calculation of a temperature field in fast reactor core using criterion equal thermo-technical reliability of subassemblies in various zones throttling taking into account change thermohydraulic characteristics of subassemblies during campaign under influence change form of core, redistribution heat generation, casual any deviation of various parameters is stated. The distribution of the statistical characteristics of a temperature field in subassemblies is calculated on subchannel method with account of an interchannel exchange and feature of influence of deformation on a temperature field in subassemblies using Monte-Carlo method. The results of the calculations show that deformation can have significant influence on a temperature mode of core. It is necessary to make thermohydraulic analysis of core during campaign at a stage of preliminary study of the projects fast reactors. (author)

Exploring the use of a deterministic adjoint flux calculation in criticality Monte Carlo simulations

International Nuclear Information System (INIS)

Jinaphanh, A.; Miss, J.; Richet, Y.; Martin, N.; Hebert, A.

2011-01-01

The paper presents a preliminary study on the use of a deterministic adjoint flux calculation to improve source convergence issues by reducing the number of iterations needed to reach the converged distribution in criticality Monte Carlo calculations. Slow source convergence in Monte Carlo eigenvalue calculations may lead to underestimate the effective multiplication factor or reaction rates. The convergence speed depends on the initial distribution and the dominance ratio. We propose using an adjoint flux estimation to modify the transition kernel according to the Importance Sampling technique. This adjoint flux is also used as the initial guess of the first generation distribution for the Monte Carlo simulation. Calculated Variance of a local estimator of current is being checked. (author)

Preliminary calculations for the CAFE project (Clean Air For Europe); Calculs preparatoires pour la strategie thematique CAFE (Clean Air For Europe)

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-09-15

The European Commission decided in 2001 an analysis program to reduce the atmospheric emissions. This report presents different limit scenari for France in 2020 (the reference scenari and the MTFR scenari, Maximum Technically Feasible Reduction), optimized scenari calculated by the RAINS model (Regional Air Pollution Information and Simulation), the costs of the scenari calculated with RAINS and the cost-benefit analysis of the strategy CAFE. From the study results, the benefits are higher than the costs, even with the most ambitious scenari. At an european level the emission reduction strategies have no effect on the employment but an impact on the Gross Domestic Product (decrease between 0,04 % and 0,12 % in function of the scenari). (A.L.B.)

Preliminary conceptual design and analysis on KALIMER reactor structures

International Nuclear Information System (INIS)

Kim, Jong Bum

1996-10-01

The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

Shielding calculations for NET

International Nuclear Information System (INIS)

Verschuur, K.A.; Hogenbirk, A.

1991-05-01

In the European Fusion Technology Programme there is only a small activity on research and development for fusion neutronics. Never-the-less, looking further than blanket design now, as ECN is getting involved in design of radiation shields for the coils and biological shields, it becomes apparent that fusion neutronics as a whole still needs substantial development. Existing exact codes for calculation of complex geometries like MCNP and DORT/TORT are put over the limits of their numerical capabilities, whilst approximate codes for complex geometries like FURNACE and MERCURE4 are put over the limits of their modelling capabilities. The main objective of this study is just to find out how far we can get with existing codes in obtaining reliable values for the radiation levels inside and outside the cryostat/shield during operation and after shut-down. Starting with a 1D torus model for preliminary parametric studies, more dimensional approximation of the torus or parts of it including the main heterogeneities should follow. Regular contacts with the NET-Team are kept, to be aware of main changes in NET design that might affect our calculation models. Work on the contract started 1 July 1990. The technical description of the contract is given. (author). 14 refs.; 4 figs.; 1 tab

Method of preliminary localization of the iris in biometric access control systems

Science.gov (United States)

Minacova, N.; Petrov, I.

2015-10-01

This paper presents a method of preliminary localization of the iris, based on the stable brightness features of the iris in images of the eye. In tests on images of eyes from publicly available databases method showed good accuracy and speed compared to existing methods preliminary localization.

A Preliminary Study on Calculation of Inter-Pebble Dancoff Factor in a Pebble Type Core

International Nuclear Information System (INIS)

Kim, Song Hyun; Kim, Hong Chul; Kim, Soon Young; Noh, Jae Man; Kim, Jong Kyung

2009-01-01

The Dancoff factor is an entering probability of the neutron escaped from specific fuel kernel to another one without the interaction with moderators. Currently, Dancoff factors are mainly evaluated from stochastic methods, hence a research on analytical method is considerably insufficient in this field. In order to analytically evaluate Dancoff factor considering double-heterogeneous effect, inter-pebble and intra-pebble Dancoff factors should be calculated, respectively. Intra-pebble Dancoff factor related with the fuel kernels in one pebble was analyzed in past study. For the evaluation of inter-pebble Dancoff factor, fuel region to region Dancoff factor (FRDF) was defined and the method to calculate the FRDF is developed in this study. The result is compared with the calculation result of the MCNP5 code

Preliminary design studies for the DESCARTES and CIDER codes

International Nuclear Information System (INIS)

Eslinger, P.W.; Miley, T.B.; Ouderkirk, S.J.; Nichols, W.E.

1992-12-01

The Hanford Environmental Dose Reconstruction (HEDR) project is developing several computer codes to model the release and transport of radionuclides into the environment. This preliminary design addresses two of these codes: Dynamic Estimates of Concentrations and Radionuclides in Terrestrial Environments (DESCARTES) and Calculation of Individual Doses from Environmental Radionuclides (CIDER). The DESCARTES code will be used to estimate the concentration of radionuclides in environmental pathways, given the output of the air transport code HATCHET. The CIDER code will use information provided by DESCARTES to estimate the dose received by an individual. This document reports on preliminary design work performed by the code development team to determine if the requirements could be met for Descartes and CIDER. The document contains three major sections: (i) a data flow diagram and discussion for DESCARTES, (ii) a data flow diagram and discussion for CIDER, and (iii) a series of brief statements regarding the design approach required to address each code requirement

Vestibule and Cask Preparation Mechanical Handling Calculation

International Nuclear Information System (INIS)

Ambre, N.

2004-01-01

The scope of this document is to develop the size, operational envelopes, and major requirements of the equipment to be used in the vestibule, cask preparation area, and the crane maintenance area of the Fuel Handling Facility. This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAIC Company L.L.C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC--28-01R W12101'' (Ref. 167124). This correspondence was appended by further correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC--28-01R W12101; TDL No. 04-024'' (Ref. 16875 1). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process

Calculation study of the WWER-440 fuel performance for extended burnup

International Nuclear Information System (INIS)

Kujal, J.; Pazdera, F.; Barta, O.

1984-01-01

The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

Space Elevators Preliminary Architectural View

Science.gov (United States)

Pullum, L.; Swan, P. A.

Space Systems Architecture has been expanded into a process by the US Department of Defense for their large scale systems of systems development programs. This paper uses the steps in the process to establishes a framework for Space Elevator systems to be developed and provides a methodology to manage complexity. This new approach to developing a family of systems is based upon three architectural views: Operational View OV), Systems View (SV), and Technical Standards View (TV). The top level view of the process establishes the stages for the development of the first Space Elevator and is called Architectural View - 1, Overview and Summary. This paper will show the guidelines and steps of the process while focusing upon components of the Space Elevator Preliminary Architecture View. This Preliminary Architecture View is presented as a draft starting point for the Space Elevator Project.

Evaluation of a new software tool for the automatic volume calculation of hepatic tumors. First results

International Nuclear Information System (INIS)

Meier, S.; Mildenberger, P.; Pitton, M.; Thelen, M.; Schenk, A.; Bourquain, H.

2004-01-01

Purpose: computed tomography has become the preferred method in detecting liver carcinomas. The introduction of spiral CT added volumetric assessment of intrahepatic tumors, which was unattainable in the clinical routine with incremental CT due to complex planimetric revisions and excessive computing time. In an ongoing clinical study, a new software tool was tested for the automatic detection of tumor volume and the time needed for this procedure. Materials and methods: we analyzed patients suffering from hepatocellular carcinoma (HCC). All patients underwent treatment with repeated transcatheter chemoembolization of the hepatic arteria. The volumes of the HCC lesions detected in CT were measured with the new software tool in HepaVison (MeVis, Germany). The results were compared with manual planimetric calculation of the volume performed by three independent radiologists. Results: our first results in 16 patients show a correlation between the automatically and the manually calculated volumes (up to a difference of 2 ml) of 96.8%. While the manual method of analyzing the volume of a lesion requires 2.5 minutes on average, the automatic method merely requires about 30 seconds of user interaction time. Conclusion: These preliminary results show a good correlation between automatic and manual calculations of the tumor volume. The new software tool requires less time for accurate determination of the tumor volume and can be applied in the daily clinical routine. (orig.) [de

Original Article PRELIMINARY BIOAUTOGRAPHIC ANALYSIS OF ...

African Journals Online (AJOL)

Sierra Leone 2Department of Pharmaceutical Chemistry, Faculty of Pharmacy, ... the seeds are used in the treatment of skin infections. ... Screening with DPPH showed prominent antioxidant spots on silica at Rf 0.8, 0.5, 0.4 .... underpins conditions like rheumatoid arthritis, ..... As a follow-up to the preliminary TLC studies.

A compartment model for nuclide release calculation in the near-and far-field of a HLW repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung; Hahn, Pil Soo

2004-01-01

The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997, from which a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel is to be introduced by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as spent fuel and generic site characteristics in Korea was roughly envisaged in 2003. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near - and far - field components of the repository, even though sufficient information has not been available that much yet, but also to show a appropriate methodology by which both a generic and site - specific safety assessment could be performed for further in - depth development of Korea reference repository concept, nuclide release calculation study for various nuclide release cases is mandatory. To this end a similar study done and yet limited for the near - field release case has been extended to the case including far - field system by introducing some more geosphere compartments. Advective and longitudinal dispersive nuclide transports along the fracture with matrix diffusion as well as several retention mechanisms and nuclide ingrowth has been added

Presentation and qualification of criticality calculation in fuel element storage

International Nuclear Information System (INIS)

Ermumcu, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

Faced with the growing size of criticality calculation requests a fast and slightly conservative method has been perfected for evaluating the effective multiplication constant of sites containing PWR type elements. This method is based on the use of the DOT 3.5 code which requires a bidimensional modelisation of the geometry of the problem and the placing into groups of the macroscopic cross sections of the various materials. This preliminary work is effected by various APOLLO calculations. This diagram is qualified by comparison with the results obtained by the Monte Carlo TRIPOLI code. Comparing the values obtained by MORET and APOLLO-DOT for the criticality of transport flask end in good agreement. For the parametric studies, a large number of calculations can be necessary, and analytical methods cost little for simple geometries. This diagram can be used for studying small transport flasks but it is particularly advantageous for storages [fr

Preliminary calculations on the cooling rate of the Renca batholit, Sierra de San Luis, Argentina

International Nuclear Information System (INIS)

Lopez de Luchi, M.G.; Ostera, H.A.; Linares, E; Rosello, E.A

2001-01-01

Cooling rates can be used to constrain the unroofing history of plutonic-metamorphic system. Geocronological cooling rates (Spear and Parrish, 1996) can be unravelled using age calculations on minerals that were open systems and subsequently passed through their closure temperatures (Dodson, 1973) during cooling. Several age determinations on different minerals are needed in order to accurately constrain the cooling path of a pluton (Hodges 1991, Spear and Parrish, 1996 and references therein). Isotopic open-system behaviour in minerals can be modelled as volume diffusion process (Hodges, 1991 and references therein), which depends on the cooling rate of the whole system. We present the first results on the calculation of the cooling rate of the Renca batholith on the basis of the combination of both thermometric calculations and available crystallization and cooling ages (au)

An extension of the fenske-hall LCAO method for approximate calculations of inner-shell binding energies of molecules

Science.gov (United States)

Zwanziger, Ch.; Reinhold, J.

1980-02-01

The approximate LCAO MO method of Fenske and Hall has been extended to an all-election method allowing the calculation of inner-shell binding energies of molecules and their chemical shifts. Preliminary results are given.

Calculations of core-excited states in Li

International Nuclear Information System (INIS)

Verbockhaven, G.; Hansen, J.E.

1999-01-01

We report on progress in the calculation of three-electron states making use of B-spline basis sets. In particular we discuss the advantages and disadvantages of using a Hartree-Fock basis (expanded in B-splines) compared to the use of hydrogenic basis states. Preliminary results are presented for the 2 S terms in Li below the 1s2s 3 S limit at 64.4 eV. The 2 S terms have been studied less extensively than other core-excited states in Li. In this particular case the choice of basis has a large influence on the quality of the results. (orig.)

Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

International Nuclear Information System (INIS)

Mishima, Kaichiro; Shibata, Toshikazu.

1982-01-01

This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

Preliminary corrosion models for BWIP [Basalt Waste Isolation Project] canister materials

International Nuclear Information System (INIS)

Fish, R.L.; Anantatmula, R.P.

1983-01-01

Waste package development for the Basalt Waste Isolation Project (BWIP) requires the generation of materials degradation data under repository relevant conditions. These data are used to develop predictive models for the behavior of each component of waste package. The component models are exercised in performance analyses to optimize the waste package design. This document presents all repository relevant canister materials corrosion data that the BWIP and others have developed to date, describes the methodology used to develop preliminary corrosion models and provides the mathematical description of the models for both low carbon steel and Fe9Cr1Mo steel. Example environment/temperature history and model application calculations are presented to aid in understanding the models. The models are preliminary in nature and will be updated as additional corrosion data become available. 6 refs., 5 tabs

Is College Affordable? Are Loans Manageable? What Do Colleges' Net Price Calculators Show? Policy Bulletin

Science.gov (United States)

Advisory Committee on Student Financial Assistance, 2012

2012-01-01

A review of net price calculators--a financial aid tool mandated by the "Higher Education Opportunity Act" of 2008--reveals that students from low-, moderate-, and middle-income families face record-level net prices at 4-year public colleges today. These net prices will translate into levels of average total loan burden far in excess of…

The SNS target station preliminary Title I shielding analyses

International Nuclear Information System (INIS)

Johnson, J.O.; Santoro, R.T.; Lillie, R.A.; Barnes, J.M.; McNeilly, G.S.

2000-01-01

The Department of Energy (DOE) has given the Spallation Neutron Source (SNS) project approval to begin Title I design of the proposed facility to be built at Oak Ridge National Laboratory (ORNL). During the conceptual design phase of the SNS project, the target station bulk-biological shield was characterized and the activation of the major targets station components was calculated. Shielding requirements were assessed with respect to weight, space, and dose-rate constraints for operating, shut-down, and accident conditions utilizing the SNS shield design criteria, DOE Order 5480.25, and requirements specified in 10 CFR 835. Since completion of the conceptual design phase, there have been major design changes to the target station as a result of the initial shielding and activation analyses, modifications brought about due to engineering concerns, and feedback from numerous external review committees. These design changes have impacted the results of the conceptual design analyses, and consequently, have required a re-investigation of the new design. Furthermore, the conceptual design shielding analysis did not address many of the details associated with the engineering design of the target station. In this paper, some of the proposed SNS target station preliminary Title I shielding design analyses will be presented. The SNS facility (with emphasis on the target station), shielding design requirements, calculational strategy, and source terms used in the analyses will be described. Preliminary results and conclusions, along with recommendations for additional analyses, will also be presented. (author)

Preliminary bathymetry; approaches to Unakwik Inlet, Alaska

Science.gov (United States)

Post, Austin

1980-01-01

A map, scale 1:20,000, shows water depths, rocks, and hazards to navigation. These data are noted on track lines run by the Research Vessel Growler in Alaskan waters, where data on navigation shown on published charts are nonexistant, preliminary, or out dated. (USGS)

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)

Thermohydraulic calculations in rectangular channels for RA-6 type reactors with transition regime

International Nuclear Information System (INIS)

Sillin, N; Vertullo, A.; Masson, V.; Hilal, R

2009-01-01

In August 2000 and within the framework of the RA-6 core conversion from high to low enrichment (20%), a preliminary analysis was performed to evaluate the maximum power that the reactor could operate with the new kernel without makeing substantial changes. This meant keeping intact, for example, the concrete shield of the pool and the nucleus inlet and outlet pipes embedded in the walls. Preliminary results indicated that for these boundary conditions a maximum power of about 3 MWt could be achieved. In August 2005 the project was resumed and new calculations performed taking as a starting point the ECBE plate fuel element(U3O8-Al). A core was developed with cooling channle widths of 2.6 mm for the control fuel elements and 2.7 mm for standard fuel elements. The thermo-hydraulic calculation puts in evidence that coolant flow into the core was in the transitional regime for the vast majority of configurations. While TERMIC code, used for thermo-hydraulic design, has been extensively tested and validated for use in research reactors under turbulent and laminar flows, this is not so for transition conditions. The transition regime is strongly dependent on conditions such as flow inlet characteristics, channel geometry, etc.. and therefore there are no reliable correlations for general use. For this reason we found it convenient to carry out experiments simulating the working conditions in order to adjust the code results with experimental data. In the present work we show the experimental results, the simulation of the experiences using the TERMIC code, and the adjustments made to the correlations used by the code so that it can be applied to the thermo-hydraulic design of the new core. [es

Waste Retrieval Sluicing System Campaign Number 3 Solids Volume Transferred Calculation

International Nuclear Information System (INIS)

CAROTHERS, K.G.

1999-01-01

Waste Retrieval Sluicing System (WRSS) operations at tank 241-C-106 began on Wednesday, November 18, 1998. The purpose of this system is to retrieve and transfer the high-heat sludge from the tank for storage in double-shell tank 241-AY-102, thereby resolving the high-heat safety issue for the tank, and to demonstrate modernized past-practice retrieval technology for single-shell tank waste. Performance Agreement (PA) TWR 1.2.2, C-106 Sluicing, was established by the Department of Energy, Office of River Protection (ORP) for achieving completion of sluicing retrieval of waste from tank 241-C-106 by September 30, 1999. This level of sludge removal is defined in the PA as either removal of approximately 72 inches of sludge or removal of 172,000 gallons of sludge (approximately 62 inches) and less than 6,000 gallons (approximately 2 inches) of sludge removal per 12 hour sluice batch for three consecutive batches. Preliminary calculations of the volume of tank 241-C-106 sludge removed as of September 29, 1999 were provided to ORP documenting completion of PA TWR 1.2.2 (Allen 1999a). The purpose of this calculation is to document the final sludge volume removed from tank 241-C-106 up through September 30, 1999. Additionally, the results of an extra batch completed October 6, 1999 is included to show the total volume of sludge removed through the end of WRSS operations. The calculation of the sludge volume transferred from the tank is guided by engineering procedure HNF-SD-WM-PROC-021, Section 15.0,Rev. 3, sub-section 4.4, ''Calculation of Sludge Transferred.''

Waste Retrieval Sluicing System Campaign Number 3 Solids Volume Transferred Calculation

International Nuclear Information System (INIS)

CAROTHERS, K.G.

1999-01-01

Waste Retrieval Sluicing System (WRSS) operations at tank 241-C-106 began on Wednesday, November 18,1998. The purpose of this system is to retrieve and transfer the high-heat sludge from the tank for storage in double-shell tank 241-AY-102, thereby resolving the high-heat safety issue for the tank, and to demonstrate modernized past-practice retrieval technology for single-shell tank waste. Performance Agreement (PA) TWR 1.2.2, C-106 Sluicing, was established by the Department of Energy, Office of River Protection (ORP) for achieving completion of sluicing retrieval of waste from tank 241-C-106 by September 30,1999. This level of sludge removal is defined in the PA as either removal of approximately 72 inches of sludge or removal of 172,000 gallons of sludge (approximately 62 inches) and less than 6,000 gallons (approximately 2 inches) of sludge removal per 12 hour sluice batch for three consecutive batches. Preliminary calculations of the volume of tank 241-C-106 sludge removed as of September 29, 1999 were provided to ORP documenting completion of PA TWR 1.2.2 (Allen 1999a). The purpose of this calculation is to document the final sludge volume removed from tank 241-C-106 up through September 30, 1999. Additionally, the results of an extra batch completed October 6, 1999 is included to show the total volume of sludge removed through the end of WRSS operations. The calculation of the sludge volume transferred from the tank is guided by engineering procedure HNF-SD-WM-PROC-021, Section 15.0,Rev. 3, sub-section 4.4, ''Calculation of Sludge Transferred.''

Preliminary assessment of the controlled release of radionuclides from waste packages containing borosilicate waste glass

International Nuclear Information System (INIS)

Strachan, D.M.; McGrail, B.P.; Apted, M.J.; Engle, D.W.; Eslinger, P.W.

1990-06-01

The purpose of this report is to provide a preliminary assessment of the release-rate for an engineered barriers subsystem (EBS) containing waste packages of defense high-level waste borosilicate glass at geochemical and hydrological conditions similar to the those at Yucca Mountain. The relationship between the proposed Waste Acceptance Preliminary Specifications (WAPS) test of glass- dissolution rate and compliance with the NRC's release-rate criterion is also evaluated. Calculations are reported for three hierarchical levels: EBS analysis, waste-package analysis, and waste-glass analysis. The following conclusions identify those factors that most acutely affect the magnitude of, or uncertainty in, release-rate performance

Preliminary Cost Estimates for Nuclear Hydrogen Production: HTSE System

International Nuclear Information System (INIS)

Yang, K. J.; Lee, K. Y.; Lee, T. H.

2008-01-01

KAERI is now focusing on the research and development of the key technologies required for the design and realization of a nuclear hydrogen production system. As a preliminary study of cost estimates for nuclear hydrogen systems, the hydrogen production costs of the nuclear energy sources benchmarking GTMHR and PBMR are estimated in the necessary input data on a Korean specific basis. G4-ECONS was appropriately modified to calculate the cost for hydrogen production of HTSE (High Temperature Steam Electrolysis) process with VHTR (Very High Temperature nuclear Reactor) as a thermal energy source. The estimated costs presented in this paper show that hydrogen production by the VHTR could be competitive with current techniques of hydrogen production from fossil fuels if CO 2 capture and sequestration is required. Nuclear production of hydrogen would allow large-scale production of hydrogen at economic prices while avoiding the release of CO 2 . Nuclear production of hydrogen could thus become the enabling technology for the hydrogen economy. The major factors that would affect the cost of hydrogen were also discussed

Calculated and experimental substantiation of the thermal method for non-destructive testing of fuel elements

International Nuclear Information System (INIS)

Maksimov, N.M.; Soldatenko, V.A.; Petrovichev, V.I.; Salimov, S.E.; Aleksandrov, K.A.; Kurov, D.A.

1985-01-01

The main systems and methods of thermal testing, their potentialities and advantages, thermal irradiation photodetectors are described. Possible fields of application of thermal testing in nuclear engineering are discussed. Calculations of the fuel element nonstationary temperature field in the three-dimensional geometry in the presence of such an effect as fuel exfaliation from cladding are presented. The developed method and equipment for fuel element thermal testing are described. Preliminary experimental data being in agreement with the calculated ones and opening the prospects for flaw detecting are presened

Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors

International Nuclear Information System (INIS)

Chambers, R.; Laats, E.T.

1981-01-01

A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper

CONTAIN calculations

International Nuclear Information System (INIS)

Scholtyssek, W.

1995-01-01

In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

MARS input data for steady-state calculation of ATLAS

International Nuclear Information System (INIS)

Park, Hyun Sik; Euh, D. J.; Choi, K. Y.; Kwon, T. S.; Jeong, J. J.; Baek, W. P.

2004-12-01

An integral effect test loop for Pressurized Water Reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), is under construction by Thermal-Hydraulics Safety Research Division in Korea Atomic Energy Research Institute (KAERI). This report includes calculation sheets of the input for the best-estimate system analysis code, the MARS code, based on the ongoing design features of ATLAS. The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400. The contents of this report are divided into three parts: (1) core and reactor vessel, (2) steam generator and steam line, and (3) primary piping, pressurizer and reactor coolant pump. The steady-state analysis for the ATLAS facility will be performed based on these calculation sheets, and its results will be applied to the detailed design of ATLAS. Additionally, the calculation results will contribute to getting optimum test conditions and preliminary operational test conditions for the steady-state and transient experiments

Results of calculation of WWER-440 fuel rods (Kol`skaya-3 NPP) at high burnup

Energy Technology Data Exchange (ETDEWEB)

Scheglov, A; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Panin, M; Pitkin, Yu [Kol` skaya NPP, (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation)

1994-12-31

Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol`skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300{sup o} C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs.

Preliminary Monthly Climatological Summaries

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Preliminary Local Climatological Data, recorded since 1970 on Weather Burean Form 1030 and then National Weather Service Form F-6. The preliminary climate data pages...

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

Science.gov (United States)

Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno

2018-01-01

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

Directory of Open Access Journals (Sweden)

Clamens Olivier

2018-01-01

Full Text Available CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Conclusion of the Preliminary Safety report for the LILW Repository on Trgovska Gora

International Nuclear Information System (INIS)

Lokner, V.; Levanat, I.; Schaller, A.; Kucar-Dragicevic, S.; Cerskov Klika, M.; Subasic, D.

2002-01-01

For more than a decade, APO d.o.o. has been engaged in preparations which might lead to establishment of a radioactive waste repository on Trgovska Gora, suitable for disposal of low and intermediate level waste (LILW) from the nuclear power plant Krsko. A recent product of theses activities is the preliminary safety assessment report (PSAR) for the proposed repository. In addition to an extensive overview of the repository project status, this preliminary SAR describes how the safety assessment methodology is used to demonstrate that a LILW facility will comply with radiological protection and safety requirements after the repository closure. LILW repository is designed to isolate waste from the environment for a couple hundred years in a reasonably efficient manner. It is generally not practicable to grant full waste containment throughout that period, because it suffices to demonstrate that radionuclide release and migration will remain below acceptable levels, which is achieved through safety assessment scenarios, modeling and calculations. However, with very limited repository specific data, safety assessment can only produce a conservative estimate of the upper bounds of potential exposures the repository could inflict. This PSAR arrives at such estimates in two different ways: (a) by simple bounding calculations and (b) through more sophisticated modeling and application of dedicated computer codes, but with similar conservative assumptions. Both approaches conservatively estimate that the highest potential dose to a nearby resident cannot significantly exceed the dose constraint of 0.2 mSv per year. Only in case of inadvertent intrusion into the near-surface disposal vault, much higher doses might be inflicted immediately after the planned institutional control of 250 years expires, but that can be prevented by a longer control period. Despite the preliminary and bounding style of the calculations, the PSAR has identified most important assumptions and

Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

Energy Technology Data Exchange (ETDEWEB)

Zhao, Zhou, E-mail: zhaozhou@swip.ac.cn; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

2016-02-15

Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li{sub 4}SiO{sub 4} pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

International Nuclear Information System (INIS)

Zhao, Zhou; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

2016-01-01

Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li_4SiO_4 pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

Preliminary report on the experiment performed in MARIUS reactor loaded with teledial fuel

Energy Technology Data Exchange (ETDEWEB)

Estiot, J C; Morier, F

1972-06-15

The experimental work described in this paper is part of a collaborative programme agreed between CEA and the Dragon Project. The aim of the programme is the measurement of the relative conversion ratio in a reactor loaded with Teledial fuel elements. The results will allow us to check our calculational methods and assumptions upon which the calculations are based, in the case of a teledial core, which represents a very complicated geometry, specially, due to the presence of the U238 with its resonance. The programme of experiments described in the paper have been completed. Some preliminary results are presented in the second part of this report (Part 2).

Preliminary estimation of the dose rates of the operation room of the RPR radioisotope cell

International Nuclear Information System (INIS)

Rocha, A.C.S.; Silva, J.J.G.; Pina, J.L.S. de; Fajardo, P.W.

1986-07-01

During the preliminary studies, about the installations layout of a radioisotope production reactor, the possibility of construction of a radioisotope cell at the reactor building has been investigated. The decisions about that construction has considered mainly the level of the radiation dose over the cell operator. The dose rate has been calculated based on: neutron flux and gamma radiation from fission products and activation materials inside the reactor; volatile fission products such as noble gases and iodides; tritium form ternary fission. The objective was calculate the radiation dose over the cell operator during a journey of 8 hours of work per day. For those calculations some data have been obtained from the Angra-3 reactor. (author)

Preliminary characterization of slow growing rhizobial strains ...

African Journals Online (AJOL)

In this paper, we did some preliminary characterization of six slow growing rhizobial strains, isolated from Retama monosperma (L.) Boiss. root nodules sampled from 3 sites along the coast of Oran (CapeFalcon, Bousfer and MersElHadjadj) in Northwestern Algeria. Results of this study showed that all strains had a very ...

77 FR 73417 - Diamond Sawblades and Parts Thereof From the People's Republic of China: Preliminary Results of...

Science.gov (United States)

2012-12-10

... Thereof From the People's Republic of China: Preliminary Results of Antidumping Duty Administrative Review... Republic of China (``PRC''). The period of review (``POR'') is November 1, 2010 through October 31, 2011... have been calculated in accordance with section 772 of the Act. Because the PRC is a nonmarket economy...

Preliminary bathymetry; Ester Passage to Eaglek Island, Alaska

Science.gov (United States)

Post, Austin

1980-01-01

A map, scale 1:20,000, shows water depths, rocks, and hazards to navigation. These data are noted on track lines run by the Research Vessel Growler in Alaskan waters, where data on navigation shown on published charts are nonexistant, preliminary, or out dated. (USGS)

Preliminary design of a borax internal core-catcher for a gas cooled fast reactor

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Schumacher, G.

1976-09-01

Preliminary thermal calculations show that a core-catcher appears to be feasible, which is able to cope with the complete meltdown of the core and blankets of a 1,000 MWe GCFR. This core-catcher is based on borax (Na 2 B 4 O 7 ) as dissolving material of the oxide fuel and of the fission products occuring in oxide form. The borax is contained in steel boxes forming a 2.1 meter thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel, just underneath the reactor core. The fission products are dispersed in the pool formed by the liquid borax. The heat power density in the pool is conveniently reduced and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system. (orig.) [de

Preliminary risk benefit assessment for nuclear waste disposal in space

Science.gov (United States)

Rice, E. E.; Denning, R. S.; Friedlander, A. L.; Priest, C. C.

1982-01-01

This paper describes the recent work of the authors on the evaluation of health risk benefits of space disposal of nuclear waste. The paper describes a risk model approach that has been developed to estimate the non-recoverable, cumulative, expected radionuclide release to the earth's biosphere for different options of nuclear waste disposal in space. Risk estimates for the disposal of nuclear waste in a mined geologic repository and the short- and long-term risk estimates for space disposal were developed. The results showed that the preliminary estimates of space disposal risks are low, even with the estimated uncertainty bounds. If calculated release risks for mined geologic repositories remain as low as given by the U.S. DOE, and U.S. EPA requirements continue to be met, then no additional space disposal study effort in the U.S. is warranted at this time. If risks perceived by the public are significant in the acceptance of mined geologic repositories, then consideration of space disposal as a complement to the mined geologic repository is warranted.

Three-dimensional solution structure of a DNA duplex containing the BclI restriction sequence: Two-dimensional NMR studies, distance geometry calculations, and refinement by back-calculation of the NOESY spectrum

International Nuclear Information System (INIS)

Banks, K.M.; Hare, D.R.; Reid, B.R.

1989-01-01

A three-dimensional solution structure for the self-complementary dodecanucleotide [(d-GCCTGATCAGGC)] 2 has been determined by distance geometry with further refinements being performed after back-calculation of the NOESY spectrum. This DNA dodecamer contains the hexamer [d(TGATCA)] 2 recognized and cut by the restriction endonuclease BclI, and its structure was determined in hopes of obtaining a better understanding of the sequence-specific interactions which occur between proteins and DNA. Preliminary examination of the structure indicates the structure is underwound with respect to idealized B-form DNA though some of the local structural parameters (glycosyl torsion angle and pseudorotation angle) suggest a B-family type of structure is present. This research demonstrates the requirements (resonance assignments, interproton distance measurements, distance geometry calculations, and NOESY spectra back-calculation) to generate experimentally self-consistent solution structures for short DNA sequences

Density-functional calculations of the surface tension of liquid Al and Na

Science.gov (United States)

Stroud, D.; Grimson, M. J.

1984-01-01

Calculations of the surface tensions of liquid Al and Na are described using the full ionic density functional formalism of Wood and Stroud (1983). Surface tensions are in good agreement with experiment in both cases, with results substantially better for Al than those found previously in the gradient approximation. Preliminary minimization with respect to surface profile leads to an oscillatory profile superimposed on a nearly steplike ionic density disribution; the oscillations have a wavellength of about a hardsphere diameter.

BIPS-FS preliminary design, miscellaneous notes

International Nuclear Information System (INIS)

1976-01-01

A compendium of flight system preliminary design internal memos and progress report extracts for the Brayton Isotope Power System Preliminary Design Review to be held July 20, 21, and 22, 1975 is presented. The purpose is to bring together those published items which relate only to the preliminary design of the Flight System, Task 2 of Phase I. This preliminary design effort was required to ensure that the Ground Demonstration System will represent the Flight System as closely as possible

75 FR 51004 - Drill Pipe From the People's Republic of China: Preliminary Determination of Sales at Less Than...

Science.gov (United States)

2010-08-18

... calculation of production costs invalid under the Department's normal methodologies. See, e.g., Preliminary..., Ukraine, and Peru are countries comparable to the PRC in terms of economic development. See April 20, 2010... circumstances, on the NME producer's factors of production (``FOPs'') valued in a surrogate market-economy...

45 CFR 150.217 - Preliminary determination.

Science.gov (United States)

2010-10-01

... 45 Public Welfare 1 2010-10-01 2010-10-01 false Preliminary determination. 150.217 Section 150.217... Are Failing To Substantially Enforce HIPAA Requirements § 150.217 Preliminary determination. If, at... designees). (b) Notifies the State of CMS's preliminary determination that the State has failed to...

Preliminary analysis of engineered barrieer performances in geological disposal of high level waste

International Nuclear Information System (INIS)

Ohe, Toshiaki; Maki, Yasuo; Tanaka, Hiroshi; Kawanishi, Motoi.

1988-01-01

This report represents preliminary results of safety analysis of a engineered barrier system in geological disposal of high level radioactive waste. Three well-known computer codes; ORIGEN 2, TRUMP, and SWIFT were used in the simulation. Main conceptual design of the repository was almost identical to that of SKB in Sweden and NAGRA in Switzerland; the engineered barrier conasists glass solidified waste, steel overpack, and compacted bentonite. Two different underground formations are considered; granite and neogene sedimentary rock, which are typically found in Japan. We first determined the repository configuration, particularly the space between disposal pitts. The ORIGEN 2 was used to estimate heat generation in the waste glass reprocessed at 4 years after removal from PWR. Then, temperature distribution was calculated by the TRUMP. The results of two or three dimensional calculation indicated that the pit interval should be kept more than 5 m in the case of granite formation at 500 m depth, according to the temperature criteria in the bentonite layer ( 90 Sr, 241 Am, 239 Pu, and 237 Np were chosen in one or two dimensional calculations. For both cases of steady release and instanteneous release, the maximum concentration in the pore water at the boundary between bentonite and surrounding rock had the following order; 237 Np> 239 Pu> 90 Sr> 241 Am. Sensitivity analysis showed that the order mainly due to the different adsorption characteristics of the nuclides in bentonite layer. (author)

Preliminary results for the k0-INAA methodology implementation at the Neutron Activation Analysis Laboratory, LAN-IPEN, using k0-IAEA software

International Nuclear Information System (INIS)

Mariano, Davi B.; Figueiredo, Ana Maria G.; Semmler, Renato

2009-01-01

The present paper presents the preliminary results obtained in the implementation of the k 0 standardization method at the Neutron Activation Laboratory (LAN) at IPEN, Sao Paulo, Brazil, using the program k 0 -IAEA, provided by The International Atomic Energy Agency (IAEA). This method is an important alternative for the comparative neutron activation analysis, which has been used for several years at LAN-IPEN. This quasiabsolute standardization method presents a great advantage with relation to the comparative method, since it does not require the preparation of accurate individual standards for each analysed element, which is very laborious and time-consuming. The k 0 method allows the determination of almost all elements whose gammaray peaks are present in the gamma spectrum. The analysis of gamma-ray spectra and the calculation of concentration are performed by the k 0 software, thus the analysis time is shortened: the time spent to calculate, for instance, the concentration of 25 elements in 10 samples takes about 5 minutes.The efficiency curve of one of the gamma-ray spectrometers used at LAN was determined by measuring calibrated radioactive sources at the usually utilised counting geometries. The parameters α and f were determined by irradiating a Certified Nuclear Reference Material IRMM-530R Al-0,1% Au alloys and high purity zirconium comparators at the IEA-R1 nuclear reactor of IPEN. In order to evaluate the efficiency of the methodology, the geological reference material basalt JB-1 (GSJ) was analysed. The preliminary results obtained showed promising results in spite of some discrepancies of the data in comparison to certified values. These preliminary results indicate that some improvements in the parameters required for the use of the k 0 -IAEA software should be made so that the k 0 - NAA software can be completely successful. (author)

A flexible Monte Carlo tool for patient or phantom specific calculations: comparison with preliminary validation measurements

Science.gov (United States)

Davidson, S.; Cui, J.; Followill, D.; Ibbott, G.; Deasy, J.

2008-02-01

The Dose Planning Method (DPM) is one of several 'fast' Monte Carlo (MC) computer codes designed to produce an accurate dose calculation for advanced clinical applications. We have developed a flexible machine modeling process and validation tests for open-field and IMRT calculations. To complement the DPM code, a practical and versatile source model has been developed, whose parameters are derived from a standard set of planning system commissioning measurements. The primary photon spectrum and the spectrum resulting from the flattening filter are modeled by a Fatigue function, cut-off by a multiplying Fermi function, which effectively regularizes the difficult energy spectrum determination process. Commonly-used functions are applied to represent the off-axis softening, increasing primary fluence with increasing angle ('the horn effect'), and electron contamination. The patient dependent aspect of the MC dose calculation utilizes the multi-leaf collimator (MLC) leaf sequence file exported from the treatment planning system DICOM output, coupled with the source model, to derive the particle transport. This model has been commissioned for Varian 2100C 6 MV and 18 MV photon beams using percent depth dose, dose profiles, and output factors. A 3-D conformal plan and an IMRT plan delivered to an anthropomorphic thorax phantom were used to benchmark the model. The calculated results were compared to Pinnacle v7.6c results and measurements made using radiochromic film and thermoluminescent detectors (TLD).

Preliminary Hazards Analysis Plasma Hearth Process

International Nuclear Information System (INIS)

Aycock, M.; Coordes, D.; Russell, J.; TenBrook, W.; Yimbo, P.

1993-11-01

This Preliminary Hazards Analysis (PHA) for the Plasma Hearth Process (PHP) follows the requirements of United States Department of Energy (DOE) Order 5480.23 (DOE, 1992a), DOE Order 5480.21 (DOE, 1991d), DOE Order 5480.22 (DOE, 1992c), DOE Order 5481.1B (DOE, 1986), and the guidance provided in DOE Standards DOE-STD-1027-92 (DOE, 1992b). Consideration is given to ft proposed regulations published as 10 CFR 830 (DOE, 1993) and DOE Safety Guide SG 830.110 (DOE, 1992b). The purpose of performing a PRA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PRA then is followed by a Preliminary Safety Analysis Report (PSAR) performed during Title I and II design. This PSAR then leads to performance of the Final Safety Analysis Report performed during construction, testing, and acceptance and completed before routine operation. Radiological assessments indicate that a PHP facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous material assessments indicate that a PHP facility will be a Low Hazard facility having no significant impacts either onsite or offsite to personnel and the environment

Georgetown University Integrated Community Energy System (GU-ICES). Phase III, Stage II. Preliminary design. Final report

Energy Technology Data Exchange (ETDEWEB)

None

1980-11-01

Results are presented for two elements in the Georgetown University ICES program - the installation of a 2500-kW backpressure steam-turbine generator within a new extension to the heating and cooling plant (cogeneration) and the provision of four additional ash silos for the university's atmospheric fluidized-bed boiler plant (added storage scheme). The preliminary design and supporting documentation for the work items and architectural drawings are presented. Section 1 discusses the basis for the report, followed by sections on: feasibility analysis update; preliminary design documents; instrumentation and testing; revised work management plan; and appendices including outline constructions, turbine-generator prepurchase specification, design calculations, cost estimates, and Potomac Electric Company data. (MCW)

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

DECOVALEX-THMC Project. Task A. Influence of near field coupled THM phenomena on the performance of a spent fuel repository. Report of Task A1: Preliminary scoping calculations

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Son (ed.) [Canadian Nuclear Safety Commission (Canada); Lanru Jing (ed.) [Royal Institute of Technology, Stockholm (Sweden); Boergesson, Lennart [Clay Technology AB, Lund (Sweden); Chijimatzu, Masakazu [Hazama Corporation (Japan); Jussila, Petri [Helsinki Univ. of Technology, Helsinki (Finland); Rutqvist, Jonny [Lawrence Berkeley National Laboratory CA (United States)

2007-02-15

The DECOVALEX-THMC project is an ongoing international co-operative project that was stared in 2004 to support the development of mathematical models of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological media for siting potential nuclear fuel waste repositories. The general objective is to characterise and evaluate the coupled THMC processes in the near field and far field of a geological repository and to assess their impact on performance assessment: - during the three phases of repository development: excavation phase, operation phase and post-closure phase; - for three different rocks types: crystalline, argillaceous and tuff; - with specific focus on the issues of: Excavation Damaged Zone (EDZ), permanent property changes of rock masses, and glaciation and permafrost phenomena. The project involves a large number of research teams supported by radioactive waste management agencies or governmental regulatory bodies in Canada, China, Finland, France, Germany, Japan, Sweden and USA, who conducted advanced studies and numerical modelling of coupled THMC processes under five tasks. This report presents the definition of the first phase, Task A-1, of the Task A of the project. The task is a working example of how interaction between THMC modelling and SA analysis could be performed. Starting with the technical definition of the Task A, the report presents the results of preliminary THM calculations with a purpose of an initial appreciation of the phenomena and material properties that must be better understood in subsequent phases. Many simplifications and assumptions were introduced and the results should be considered under these assumptions. Based on the evaluation of the multiple teams' results, a few points of concern were identified that may guide the successive phases of Task A studies: 1. The predicted maximum total stress in the MX-80 bentonite could slightly exceed the 15 MPa design pressure for the

DECOVALEX-THMC Project. Task A. Influence of near field coupled THM phenomena on the performance of a spent fuel repository. Report of Task A1: Preliminary scoping calculations

International Nuclear Information System (INIS)

Nguyen, Son; Lanru Jing; Boergesson, Lennart; Chijimatzu, Masakazu; Jussila, Petri; Rutqvist, Jonny

2007-02-01

The DECOVALEX-THMC project is an ongoing international co-operative project that was stared in 2004 to support the development of mathematical models of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological media for siting potential nuclear fuel waste repositories. The general objective is to characterise and evaluate the coupled THMC processes in the near field and far field of a geological repository and to assess their impact on performance assessment: - during the three phases of repository development: excavation phase, operation phase and post-closure phase; - for three different rocks types: crystalline, argillaceous and tuff; - with specific focus on the issues of: Excavation Damaged Zone (EDZ), permanent property changes of rock masses, and glaciation and permafrost phenomena. The project involves a large number of research teams supported by radioactive waste management agencies or governmental regulatory bodies in Canada, China, Finland, France, Germany, Japan, Sweden and USA, who conducted advanced studies and numerical modelling of coupled THMC processes under five tasks. This report presents the definition of the first phase, Task A-1, of the Task A of the project. The task is a working example of how interaction between THMC modelling and SA analysis could be performed. Starting with the technical definition of the Task A, the report presents the results of preliminary THM calculations with a purpose of an initial appreciation of the phenomena and material properties that must be better understood in subsequent phases. Many simplifications and assumptions were introduced and the results should be considered under these assumptions. Based on the evaluation of the multiple teams' results, a few points of concern were identified that may guide the successive phases of Task A studies: 1. The predicted maximum total stress in the MX-80 bentonite could slightly exceed the 15 MPa design pressure for the container

Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

International Nuclear Information System (INIS)

Fletcher, C.D.

1986-01-01

The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

Magnetic field calculations for the technical proposal of the TESLA spectrometer magnet

International Nuclear Information System (INIS)

Morozov, N.A.; Schreiber, H.J.

2003-01-01

The TESLA electron-positron linear collider is under consideration at DESY (Hamburg). The realization of the physical program at this collider requires the knowledge of the beam energy of both beams (e + and e - ) with a precision of ΔE/E ≤ 10 -4 . The magnetic spectrometer was proposed as an energy measuring device. The report describes calculations for the preliminary conceptual design of this type of the spectrometer. The 2D calculations of the magnetic field for the spectrometer magnet have been performed by POISSON SUPERFISH computer code. The basic technical parameters of the magnet have been determined. These data will serve as a basis for the technical design of the spectrometer magnet and discuss its integration in the spectrometer

Exploratory shaft facility preliminary designs - Permian Basin

International Nuclear Information System (INIS)

1983-09-01

The purpose of the Preliminary Design Report, Permian Basin, is to provide a description of the preliminary design for an Exploratory Shaft Facility in the Permian Basin, Texas. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers are included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description, and Construction Cost Estimate. 30 references, 13 tables

Preliminary design of the cold neutron source for the Centro Atomico Bariloche Electron LINAC Facility. I. Solid benzene as moderating material

International Nuclear Information System (INIS)

Torres, Lourdes; Granada, Jose R.

2004-01-01

We present the results of preliminary calculations performed with the code MCNP-4C relative to the neutron field behavior within the moderator for the CAB-LINAC cold neutron source, using benzene at 89 K as moderating material. Throughout the design calculations nuclear data libraries previously generated and validated were used. The optimum dimensions for a slab and a grid moderator were calculated, with and without a pre moderator, from the point of view of neutron production and the time-width of the neutron pulse. (author)

A preliminary bending fatigue spectrum for steel monostrand cables

DEFF Research Database (Denmark)

Winkler, Jan; Fischer, Gregor; Georgakis, Christos T.

2011-01-01

This paper presents the results of the experimental study on the bending fatigue resistance of high-strength steel monostrand cables. From the conducted fatigue tests in the high-stress, low-cycle region, a preliminary bending fatigue spectrum is derived for the estimation of monostrand cable...... service life expectancy. The presented preliminary bending fatigue spectrum of high-strength monostrands is currently unavailable in the published literature. The presented results provide relevant information on the bending mechanism and fatigue characteristics of monostrand steel cables in tension...... and flexure and show that localized cable bending has a pronounced influence on the fatigue resistance of cables under dynamic excitations....

Effect of preliminary plastic deformation on low temperature strength of carbon steels

International Nuclear Information System (INIS)

Gur'ev, A.V.; Alkhimenkov, T.B.

1979-01-01

Considered is the effect of preliminary plastic deformation on the following low-temperature strength (at -196 deg C) of structural carbon steels at the room temperature. The study of regularities of microheterogenetic deformations by alloy structure elements at room and low temperatures shows that the transition on low -temperature loading is built on the base of inheritance of the general mechanism of plastic deformation, which took place at preliminary deformation; in this effect the ''memory'' of metal to the history of loading is shown. It is established that physical strengthening (cold hardening), received by the metal during preliminary loading at the room temperature is put over the strengthening connected only with decrease of test temperature

Analysis of noncondensable effect during small break transient in VVER-440 geometry with CATHARE V1.3L. Preliminary results

International Nuclear Information System (INIS)

Sarrette, C.

1996-11-01

The report presents a study of the transport and dissolution-release of non-condensable gas into the fluid of the primary loop for the VVER-440 geometry. The analysis has been done using a new model developed for the CATHARE thermal hydraulic code. Results are presented, obtained from calculations of small break loss-of-coolant (SBLOCA) accidents for the Loviisa nuclear power plant (NPP) geometry. The influence of nitrogen dissolved in the water of the accumulators of the emergency core coolant system (ECCS) on natural circulation is discussed. Possibilities of formation of nitrogen bubbles in the main vessels upper plenum, top of the downcomer, steam generators collectors, and upper structures of RCP's are investigated. First results show that there is potentiality for interruption, mainly due to the presence of nitrogen in the top of the downcomer and the upper parts of the RCP's. These preliminary results should be confirmed by carrying out calculations now prematurely stopped for numerical reasons. (8 refs.)

Mixed first- and second-order transport method using domain decomposition techniques for reactor core calculations

International Nuclear Information System (INIS)

Girardi, E.; Ruggieri, J.M.

2003-01-01

The aim of this paper is to present the last developments made on a domain decomposition method applied to reactor core calculations. In this method, two kind of balance equation with two different numerical methods dealing with two different unknowns are coupled. In the first part the two balance transport equations (first order and second order one) are presented with the corresponding following numerical methods: Variational Nodal Method and Discrete Ordinate Nodal Method. In the second part, the Multi-Method/Multi-Domain algorithm is introduced by applying the Schwarz domain decomposition to the multigroup eigenvalue problem of the transport equation. The resulting algorithm is then provided. The projection operators used to coupled the two methods are detailed in the last part of the paper. Finally some preliminary numerical applications on benchmarks are given showing encouraging results. (authors)

Inventory calculations in sediment samples with heterogeneous plutonium activity distribution

International Nuclear Information System (INIS)

Eriksson, M.; Dahlgaard, H.

2002-01-01

A method to determine the total inventory of a heterogeneously distributed contamination of marine sediments is described. The study site is the Bylot Sound off the Thule Airbase, NW Greenland, where marine sediments became contaminated with plutonium in 1968 after a nuclear weapons accident. The calculation is based on a gamma spectrometric screening of the 241 Am concentration in 450 one-gram aliquots from 6 sediment cores. A Monte Carlo programme then simulates a probable distribution of the activity, and based on that, a total inventory is estimated by integrating a double exponential function. The present data indicate a total inventory around 3.5 kg, which is 7 times higher than earlier estimates (0.5 kg). The difference is partly explained by the inclusion of hot particles in the present calculation. A large uncertainty is connected to this estimate, and it should be regarded as preliminary. (au)

Preliminary hazards analysis -- vitrification process

International Nuclear Information System (INIS)

Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P.

1994-06-01

This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility's construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment

Preliminary hazards analysis -- vitrification process

Energy Technology Data Exchange (ETDEWEB)

Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P. [Science Applications International Corp., Pleasanton, CA (United States)

1994-06-01

This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility`s construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment.

Preliminary results from MBE-4: A four beam induction linac for heavy ion fusion research

International Nuclear Information System (INIS)

Fessenden, T.J.; Judd, D.L.; Keefe, D.; Kim, C.; Laslett, L.J.; Smith, L.; Warwick, A.I.; Warwick, P.b.A.I.

1986-01-01

Preliminary results are presented from a scaled experimental multiple beam induction linac. This experiment is part of a program of accelerator research for heavy ion fusion. It is shown that multiple beams can be accelerated without significant mutual interaction. Measurements of the longitudinal dynamics of a current-amplifying induction linac are presented and compared to calculations. Coupling of transverse and longitudinal dynamics is discussed

Preliminary results from MBE-4: a four beam induction linac for heavy ion fusion research

International Nuclear Information System (INIS)

Fessenden, T.J.; Judd, D.L.; Keefe, D.; Kim, C.; Laslett, L.J.; Smith, L.; Warwick, A.I.

1986-05-01

Preliminary results are presented from a scaled experimental multiple beam induction linac. This experiment is part of a program of accelerator research for heavy ion fusion. It is shown that multiple beams can be accelerated without significant mutual interaction. Measurements of the longitudinal dynamics of a current-amplifying induction linac are presented and compared to calculations. Coupling of transverse and longitudinal dynamics is discussed

Preliminary Design Analysis of a HGD for the NHDD Program at Korea

International Nuclear Information System (INIS)

Song, Kee Nam; Lee, H. Y.; Lee, S. B.; Kim, Y. W.

2007-01-01

Korea Atomic Energy Research Institute is in the process of carrying out a Nuclear Hydrogen Development and Demonstration (NHDD) Program by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. A coaxial double-tube Hot Gas Duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the NHDD program. Recently, a preliminary design evaluation for the hot gas duct of the NHDD program was carried out. These preliminary design activities include a decision on the geometric dimensions, a strength evaluation, an appropriate material selection, and identifying the design code for the HGD. In this study, a preliminary strength evaluation for the HGD of the NHDD program has been undertaken based on the HTR-10 design concepts. Also, a preliminary evaluation of the creep-fatigue damage for a high temperature HGD structure has been carried out according to the draft code case for Alloy 617. Preliminary strength evaluation results for the HGD showed that the geometric dimensions of the proposed HGD would be acceptable for the design requirements

Development and preliminary validation of the Opioid Abuse Risk Screener

Directory of Open Access Journals (Sweden)

Patricia Henrie-Barrus

2016-05-01

Full Text Available Prescription opioid drug abuse has reached epidemic proportions. Individuals with chronic pain represent a large population at considerable risk of abusing opioids. The Opioid Abuse Risk Screener was developed as a comprehensive self-administered measure of potential risk that includes a wide range of critical elements noted in the literature to be relevant to opioid risk. The creation, refinement, and preliminary modeling of the item pool, establishment of preliminary concurrent validity, and the determination of the factor structure are presented. The initial development and validation of the Opioid Abuse Risk Screener shows promise for effective risk stratification.

Evaluation methodology for flood damage reduction by preliminary water release from hydroelectric dams

Science.gov (United States)

Ando, T.; Kawasaki, A.; Koike, T.

2017-12-01

IPCC AR5 (2014) reported that rainfall in the middle latitudes of the Northern Hemisphere has been increasing since 1901, and it is claimed that warmer climate will increase the risk of floods. In contrast, world water demand is forecasted to exceed a sustainable supply by 40 percent by 2030. In order to avoid this expectable water shortage, securing new water resources has become an utmost challenge. However, flood risk prevention and the secure of water resources are contradictory. To solve this problem, we can use existing hydroelectric dams not only as energy resources but also for flood control. However, in case of Japan, hydroelectric dams take no responsibility for it, and benefits have not been discussed accrued by controlling flood by hydroelectric dams, namely by using preliminary water release from them. Therefore, our paper proposes methodology for assessing those benefits. This methodology has three stages as shown in Fig. 1. First, RRI model is used to model flood events, taking account of the probability of rainfall. Second, flood damage is calculated using assets in inundation areas multiplied by the inundation depths generated by that RRI model. Third, the losses stemming from preliminary water release are calculated, and adding them to flood damage, overall losses are calculated. We can evaluate the benefits by changing the volume of preliminary release. As a result, shown in Fig. 2, the use of hydroelectric dams to control flooding creates 20 billion Yen benefits, in the probability of three-day-ahead rainfall prediction of the assumed maximum rainfall in Oi River, in the Shizuoka Pref. of Japan. As the third priority in the Sendai Framework for Disaster Risk Reduction 2015-2030, `investing in disaster risk reduction for resilience - public and private investment in disaster risk prevention and reduction through structural and non-structural measures' was adopted. The accuracy of rainfall prediction is the key factor in maximizing the benefits

Practical Recommendations for the Preliminary Design Analysis of ...

African Journals Online (AJOL)

Interior-to-exterior shear ratios for equal and unequal bay frames, as well as column inflection points were obtained to serve as practical aids for preliminary analysis/design of fixed-feet multistory sway frames. Equal and unequal bay five story frames were analysed to show the validity of the recommended design ...

Relationship Between Dual Time Point FDG PET and Immunohistochemical Parameters in Preoperative Colorectal Cancer: Preliminary Study

International Nuclear Information System (INIS)

Lee, Jai Hyuen; Lee, Won Ae; Park, Seok Gun; Park, Dong Kook; Namgung, Hwan

2012-01-01

The clinical availability of 2 deoxy 2 [18F] fluoro D glucose (FDG) dual time point positron emission tomography/computerized tomography (DTPP) has been investigated in diverse oncologic fields. The aim of this preliminary study was to evaluate the relationship between various immunohistopathologic markers reflecting disease progression of colorectal cancer and parameters extracted from FDG DTPP in colorectal cancer patients. Forty seven patients with histologically confirmed colorectal cancer were analyzed in this preliminary study. FDG DTPP consisted of an early scan 1 h after FDG injection and a delayed scan 1.5 h after the early scan. Based on an analysis of FDG DTPP, we estimated the maximum standardized uptake value (SUV) of tumors on the early and delayed scans (SUV earlya nd SUV delayed, respectively). The retention index (RI) was calculated as follows: (SUV delayed- SUV early) x 100/ SUV early. The clinicopathological findings (size and T and N stages) and immunohistochemical factors [glucose transporter 1 (GLUT 1), hexokinase 2 (HK 2), p53, P504S, and β catenin] were analyzed by visual analysis. The RIs calculated from the SUVs ranged from -1.8 to 73.4 (31.8±15.5). The RIs were significantly higher in patients with high T stages (T3 and T4) than with low T stages (T1 and T2; P earlya nd SUV delayeda nd clinicopathologic parameters in this study. The RIs obtained from preoperative colorectal cancers had a significant relationship to tumor size, T staging, GLUT 1, and p53, in contrast to SUV earlyo r SUV delayed. Compared with previous reports, our results showed that RI can better predict GLUT 1 expression than HK 2 and other immunohistochemical markers. This study demonstrated that the RI might have the potential to be applied as a prognostic marked in preoperative colorectal cancer

Crystallization and preliminary X-ray analysis of Leishmania major glyoxalase I

Energy Technology Data Exchange (ETDEWEB)

Ariza, Antonio; Vickers, Tim J.; Greig, Neil; Fairlamb, Alan H.; Bond, Charles S., E-mail: c.s.bond@dundee.ac.uk [Division of Biological Chemistry and Molecular Microbiology, Wellcome Trust Biocentre, School of Life Sciences, University of Dundee, Dundee DD1 5EH,Scotland (United Kingdom)

2005-08-01

The detoxification enzyme glyoxalase I from L. major has been crystallized. Preliminary molecular-replacement calculations indicate the presence of three glyoxalase I dimers in the asymmetric unit. Glyoxalase I (GLO1) is a putative drug target for trypanosomatids, which are pathogenic protozoa that include the causative agents of leishmaniasis. Significant sequence and functional differences between Leishmania major and human GLO1 suggest that it may make a suitable template for rational inhibitor design. L. major GLO1 was crystallized in two forms: the first is extremely disordered and does not diffract, while the second, an orthorhombic form, produces diffraction to 2.0 Å. Molecular-replacement calculations indicate that there are three GLO1 dimers in the asymmetric unit, which take up a helical arrangement with their molecular dyads arranged approximately perpendicular to the c axis. Further analysis of these data are under way.

The Implementation of Cosine Similarity to Calculate Text Relevance between Two Documents

Science.gov (United States)

Gunawan, D.; Sembiring, C. A.; Budiman, M. A.

2018-03-01

Rapidly increasing number of web pages or documents leads to topic specific filtering in order to find web pages or documents efficiently. This is a preliminary research that uses cosine similarity to implement text relevance in order to find topic specific document. This research is divided into three parts. The first part is text-preprocessing. In this part, the punctuation in a document will be removed, then convert the document to lower case, implement stop word removal and then extracting the root word by using Porter Stemming algorithm. The second part is keywords weighting. Keyword weighting will be used by the next part, the text relevance calculation. Text relevance calculation will result the value between 0 and 1. The closer value to 1, then both documents are more related, vice versa.

Calculated NWIS signatures for enriched uranium metal

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.; Koehler, P.E.

1995-01-01

Nuclear Weapons Identification System (NWIS) signatures have been calculated using a Monte Carlo transport code for measurement configurations of a 252 Cf source, detectors, and a uranium metal casting. NWIS signatures consist of a wide variety of time-and frequency-analysis signatures such as the time distribution of neutrons after californium fission, the time distribution of counts in a detector after a previous count, the number of times n pulses occur in a time interval, and various frequency-analysis signatures, such as auto-power and cross-power spectral densities, coherences, and a ratio of spectral densities. This ratio is independent of detection efficiency. The analysis presented here, using the MCNP-DSP code, evaluates the applicability of this method for measurement of the 235 U content of 19-kg castings of depleted uranium and uranium with enrichments of 20, 40, 60, 80, 90, and 93.2 wt % 235 U. The dependence of the wide variety of NWIS signatures on 235 U content and possible configurations of a measurement system are presented. These preliminary calculations indicate short measurement times. Additional calculations are being performed to optimize the source-detector-moderator-casting configuration for the shortest measurement time. Although the NWIS method was developed for nuclear weapons identification, the development of a small processor now allows it to be also applied in a practical way to subcriticality measurements, nuclear fuel process monitoring and qualitative nondestructive assay of special nuclear material

Preliminary Calculation of the EROI for the Production of Gas in Russia

Directory of Open Access Journals (Sweden)

Roman Nogovitsyn

2014-09-01

Full Text Available Russia is one of the world’s largest producers of energy resources. Production of energy resources in Russia is profitable, both economically and in terms of the energy produced (as measured by EROI. At the present time, Russian oil and gas companies have a policy of energy saving, and data on energy consumption is given in annual reports. Based on these data, we can make the EROI calculation. In 2013, the EROI for the production, transportation and processing of gas for Open joint stock company (OJSC “Gazprom” was 79:1; for OJSC “NOVATEK”, 76:1; for OJSC “Yakutsk Fuel and Energy Company (YATEC”, only for production, 116:1. Currently, the situation in the oil and gas industry has come to a point when there is a need for the introduction of an energy audit.

Preliminary study of the electrolysis of aluminum sulfide in molten salts

Energy Technology Data Exchange (ETDEWEB)

Minh, N.Q.; Loutfy, R.O.; Yao, N.P.

1983-02-01

A preliminary laboratory-scale study of the electrolysis of aluminum sulfide in molten salts investigated the (1) solubility of Al/sub 2/S/sub 3/ in molten salts, (2) electrochemical behavior of Al/sub 2/S/sub 3/, and (3) electrolysis of Al/sub 2/S/sub 3/ with the determination of current efficiency as a function of current density. The solubility measurements show that MgCl/sub 2/-NaCl-KCl eutectic electrolyte at 1023 K can dissolve up to 3.3 mol % sulfide. The molar ratio of sulfur to aluminum in the eutectic is about one, which suggests that some sulfur remains undissolved, probably in the form of MgS. The experimental data and thermodynamic calculations suggest that Al/sub 2/S/sub 3/ dissolves in the eutectic to form AlS/sup +/ species in solution. Addition of AlCl/sub 3/ to the eutectic enhances the solubility of Al/sub 2/S/sub 3/; the solubility increases with increasing AlCl/sub 3/ concentration. The electrode reaction mechanism for the electrolysis of Al/sub 2/S/sub 3/ was elucidated by using linear sweep voltammetry. The cathodic reduction of aluminum-ion-containing species to aluminum proceeds by a reversible, diffusion-controlled, three-electron reaction. The anodic reaction involves the two-electron discharge of sulfide-ion-containing species, followed by the fast dimerization of sulfur atoms to S/sub 2/. Electrolysis experiments show that Al/sub 2/S/sub 3/ dissolved in molten MgCl/sub 2/-NaCl-KCl eutectic or in eutectic containing AlCl/sub 3/ can be electrolyzed to produce aluminum and sulfur. In the eutectic at 1023 K, the electrolysis can be conducted up to about 300 mA/cm/sup 2/ for the saturation solubility of Al/sub 2/S/sub 3/. Although these preliminary results are promising, additional studies are needed to elucidate many critical operating parameters before the technical potential of the electrolysis can be accurately assessed. 20 figures, 18 tables.

Core calculations of JMTR

Energy Technology Data Exchange (ETDEWEB)

Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-03-01

In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

Calculation of Rydberg interaction potentials

DEFF Research Database (Denmark)

Weber, Sebastian; Tresp, Christoph; Menke, Henri

2017-01-01

for calculating the required electric multipole moments and the inclusion of electromagnetic fields with arbitrary direction. We focus specifically on symmetry arguments and selection rules, which greatly reduce the size of the Hamiltonian matrix, enabling the direct diagonalization of the Hamiltonian up...... to higher multipole orders on a desktop computer. Finally, we present example calculations showing the relevance of the full interaction calculation to current experiments. Our software for calculating Rydberg potentials including all features discussed in this tutorial is available as open source....

Preliminary Design of a Femtosecond Oscilloscope

CERN Document Server

Gazazyan, Edmond D; Kalantaryan, Davit K; Laziev, Edouard; Margaryan, Amour

2005-01-01

The calculations on motion of electrons in a finite length electromagnetic field of linearly and circularly polarized laser beams have shown that one can use the transversal deflection of electrons on a screen at a certain distance after the interaction region for the measurement of the length and longitudinal particle distribution of femtosecond bunches. In this work the construction and preliminary parameters of various parts of a device that may be called femtosecond oscilloscope are considered. The influence of various factors, such as the energy spread and size of the electron bunches, are taken into account. For CO2 laser intensity 1016 W/cm2 and field free drift length 1m the deflection is 5.3 and 0.06 cm, while the few centimeters long interaction length between 2 mirrors requires assembling accuracy 6 mm and 1.3 micron for 20 MeV to 50 keV, respectively.

Numerical methods for calculating thermal residual stresses and hydrogen diffusion

International Nuclear Information System (INIS)

Leblond, J.B.; Devaux, J.; Dubois, D.

1983-01-01

Thermal residual stresses and hydrogen concentrations are two major factors intervening in cracking phenomena. These parameters were numerically calculated by a computer programme (TITUS) using the FEM, during the deposition of a stainless clad on a low-alloy plate. The calculation was performed with a 2-dimensional option in four successive steps: thermal transient calculation, metallurgical transient calculation (determination of the metallurgical phase proportions), elastic-plastic transient (plain strain conditions), hydrogen diffusion transient. Temperature and phase dependence of hydrogen diffusion coefficient and solubility constant. The following results were obtained: thermal calculations are very consistent with experiments at higher temperatures (due to the introduction of fusion and solidification latent heats); the consistency is not as good (by 70 degrees) for lower temperatures (below 650 degrees C); this was attributed to the non-introduction of gamma-alpha transformation latent heat. The metallurgical phase calculation indicates that the heat affected zone is almost entirely transformed into bainite after cooling down (the martensite proportion does not exceed 5%). The elastic-plastic calculations indicate that the stresses in the heat affected zone are compressive or slightly tensile; on the other hand, higher tensile stresses develop on the boundary of the heat affected zone. The transformation plasticity has a definite influence on the final stress level. The return of hydrogen to the clad during the bainitic transformation is but an incomplete phenomenon and the hydrogen concentration in the heat affected zone after cooling down to room temperature is therefore sufficient to cause cold cracking (if no heat treatment is applied). Heat treatments are efficient in lowering the hydrogen concentration. These results enable us to draw preliminary conclusions on practical means to avoid cracking. (orig.)

Preliminary Analysis of Google+'s Privacy

OpenAIRE

Mahmood, Shah; Desmedt, Yvo

2011-01-01

In this paper we provide a preliminary analysis of Google+ privacy. We identified that Google+ shares photo metadata with users who can access the photograph and discuss its potential impact on privacy. We also identified that Google+ encourages the provision of other names including maiden name, which may help criminals performing identity theft. We show that Facebook lists are a superset of Google+ circles, both functionally and logically, even though Google+ provides a better user interfac...

Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

Directory of Open Access Journals (Sweden)

Istvan Farkas

2016-08-01

Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

Daylight calculations in practice

DEFF Research Database (Denmark)

Iversen, Anne; Roy, Nicolas; Hvass, Mette

The aim of the project was to obtain a better understanding of what daylight calculations show and also to gain knowledge of how the different daylight simulation programs perform compared with each other. Experience has shown that results for the same room, obtained from two daylight simulation...... programs can give different results. This can be due to restrictions in the program itself and/or be due to the skills of the persons setting up the models. This is crucial as daylight calculations are used to document that the demands and recommendations to daylight levels outlined by building authorities....... The aim of the project was to obtain a better understanding of what daylight calculations show and also to gain knowledge of how the different daylight simulation programs perform compared with each other. Furthermore the aim was to provide knowledge of how to build up the 3D models that were...

28 CFR 2.48 - Revocation: Preliminary interview.

Science.gov (United States)

2010-07-01

... 28 Judicial Administration 1 2010-07-01 2010-07-01 false Revocation: Preliminary interview. 2.48....48 Revocation: Preliminary interview. (a) Interviewing officer. A parolee who is retaken on a warrant issued by a Commissioner shall be given a preliminary interview by an official designated by the Regional...

Reynolds number calculation and applications for curved wall jets

Directory of Open Access Journals (Sweden)

Valeriu DRAGAN

2014-09-01

Full Text Available The current paper refers to the preliminary estimation of the Reynolds number for curved wall jets. This, in turn, can be a useful tool for controlling the boundary layer mesh size near a generic curved wall which is wetted by a thin, attached jet. The method relies on analytical calculations that link the local curvature of the wall with the pressure gradient and further, the local Reynolds number. Knowing the local Reynolds number distribution, a CFD user can tailor their mesh size to more exact specifications (e.g. y+=1 for k-omega RANS models and lower the risk that the mesh is too coarse or finer than necessary.

First-principles cluster variation calculations of tetragonal-cubic transition in ZrO2

International Nuclear Information System (INIS)

Mohri, Tetsuo; Chen, Ying; Kiyokane, Naoya

2013-01-01

Highlights: ► Cluster variation method is extended to study displacive transition. ► Electronic structure total energy calculations are performed on ZrO2. ► Tetragonal-cubic transition is studied within the framework of order -disorder transition. -- Abstract: It is attempted to extend the basic idea of continuous displacement cluster variation method (CDCVM) to the study of a displacive phase transition. As a preliminary study, we focus on cubic to tetragonal transition in ZrO 2 in which oxygen atoms on the cubic lattice are displaced alternatively in the opposite direction (upward and downward) along the tetragonal axis. Within the CDCVM, displaced atoms are regarded as different atomic species, and two distinguished atoms, A-oxygen (upward shifting) and B-oxygen (downward shifting), are introduced in the description of the free energy. FLAPW electronic structure total energy calculations are performed to extract effective interaction energies among displaced oxygen atoms, and by combing them with CDCVM, the transition temperature is calculated from the first-principles

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.

A Preliminary Comparison of Motor Learning Across Different Non-invasive Brain Stimulation Paradigms Shows No Consistent Modulations

Directory of Open Access Journals (Sweden)

Virginia Lopez-Alonso

2018-04-01

Full Text Available Non-invasive brain stimulation (NIBS has been widely explored as a way to safely modulate brain activity and alter human performance for nearly three decades. Research using NIBS has grown exponentially within the last decade with promising results across a variety of clinical and healthy populations. However, recent work has shown high inter-individual variability and a lack of reproducibility of previous results. Here, we conducted a small preliminary study to explore the effects of three of the most commonly used excitatory NIBS paradigms over the primary motor cortex (M1 on motor learning (Sequential Visuomotor Isometric Pinch Force Tracking Task and secondarily relate changes in motor learning to changes in cortical excitability (MEP amplitude and SICI. We compared anodal transcranial direct current stimulation (tDCS, paired associative stimulation (PAS25, and intermittent theta burst stimulation (iTBS, along with a sham tDCS control condition. Stimulation was applied prior to motor learning. Participants (n = 28 were randomized into one of the four groups and were trained on a skilled motor task. Motor learning was measured immediately after training (online, 1 day after training (consolidation, and 1 week after training (retention. We did not find consistent differential effects on motor learning or cortical excitability across groups. Within the boundaries of our small sample sizes, we then assessed effect sizes across the NIBS groups that could help power future studies. These results, which require replication with larger samples, are consistent with previous reports of small and variable effect sizes of these interventions on motor learning.

A Preliminary Comparison of Motor Learning Across Different Non-invasive Brain Stimulation Paradigms Shows No Consistent Modulations

Science.gov (United States)

Lopez-Alonso, Virginia; Liew, Sook-Lei; Fernández del Olmo, Miguel; Cheeran, Binith; Sandrini, Marco; Abe, Mitsunari; Cohen, Leonardo G.

2018-01-01

Non-invasive brain stimulation (NIBS) has been widely explored as a way to safely modulate brain activity and alter human performance for nearly three decades. Research using NIBS has grown exponentially within the last decade with promising results across a variety of clinical and healthy populations. However, recent work has shown high inter-individual variability and a lack of reproducibility of previous results. Here, we conducted a small preliminary study to explore the effects of three of the most commonly used excitatory NIBS paradigms over the primary motor cortex (M1) on motor learning (Sequential Visuomotor Isometric Pinch Force Tracking Task) and secondarily relate changes in motor learning to changes in cortical excitability (MEP amplitude and SICI). We compared anodal transcranial direct current stimulation (tDCS), paired associative stimulation (PAS25), and intermittent theta burst stimulation (iTBS), along with a sham tDCS control condition. Stimulation was applied prior to motor learning. Participants (n = 28) were randomized into one of the four groups and were trained on a skilled motor task. Motor learning was measured immediately after training (online), 1 day after training (consolidation), and 1 week after training (retention). We did not find consistent differential effects on motor learning or cortical excitability across groups. Within the boundaries of our small sample sizes, we then assessed effect sizes across the NIBS groups that could help power future studies. These results, which require replication with larger samples, are consistent with previous reports of small and variable effect sizes of these interventions on motor learning. PMID:29740271

242Pu: Preliminary evaluation with consideration of 240Pu, and some sensitivity results

International Nuclear Information System (INIS)

Jary, J.; Lagrange, C.; Philis, C.

1978-01-01

A preliminary evaluation of 242 Pu nuclear data is presented for the neutron energy range from 10 keV to 20 MeV. The fission cross section is based upon recent experimental measurements on 242 Pu. The remaining cross sections have been calculated using various nuclear models with parameters obtained mainly by both fits on 240 Pu experimental data and general reflexions on the actinides. Particular care has been taken of the direct interactions. The laws of secondary neutron energy spectra and the average number of neutrons produced per fission have been evaluated. The results have been placed in ENDF/BIV format and combined with the low energy region of ENDF/BIV MAT = 1161 data to make complete the evaluation over the whole energy range 10 -5 eV - 20 MeV. Finally, the sensitivities of some of these nuclear data available for reactor calculations are given in terms of the variation of the calculated critical masses

Spatial Heterodyne Observation of Water (SHOW) from a high altitude aircraft

Science.gov (United States)

Bourassa, A. E.; Langille, J.; Solheim, B.; Degenstein, D. A.; Letros, D.; Lloyd, N. D.; Loewen, P.

2017-12-01

The Spatial Heterodyne Observations of Water instrument (SHOW) is limb-sounding satellite prototype that is being developed in collaboration between the University of Saskatchewan, York University, the Canadian Space Agency and ABB. The SHOW instrument combines a field-widened SHS with an imaging system to observe limb-scattered sunlight in a vibrational band of water (1363 nm - 1366 nm). Currently, the instrument has been optimized for deployment on NASA's ER-2 aircraft. Flying at an altitude of 70, 000 ft the ER-2 configuration and SHOW viewing geometry provides high spatial resolution (limb-measurements of water vapor in the Upper troposphere and lower stratosphere region. During an observation campaign from July 15 - July 22, the SHOW instrument performed 10 hours of observations from the ER-2. This paper describes the SHOW measurement technique and presents the preliminary analysis and results from these flights. These observations are used to validate the SHOW measurement technique and demonstrate the sampling capabilities of the instrument.

Air-over-ground calculations of the neutron, prompt, and secondary-gamma free-in-air tissue kerma from the Hiroshima and Nagasaki devices

International Nuclear Information System (INIS)

Pace, J.V. III; Knight, J.R.; Bartine, D.E.

1982-01-01

This paper reports preliminary results of the two-dimensional discrete-ordinate, calculations for the air-over-ground transport of radiation from the Hiroshima and Nagasaki weapon devices. It was found that the gamma-ray kerma dominated the total kerma for both environments

A Disposable Tear Glucose Biosensor-Part 4: Preliminary Animal Model Study Assessing Efficacy, Safety, and Feasibility.

Science.gov (United States)

La Belle, Jeffrey T; Engelschall, Erica; Lan, Kenneth; Shah, Pankti; Saez, Neil; Maxwell, Stephanie; Adamson, Teagan; Abou-Eid, Michelle; McAferty, Kenyon; Patel, Dharmendra R; Cook, Curtiss B

2014-01-01

A prototype tear glucose (TG) sensor was tested in New Zealand white rabbits to assess eye irritation, blood glucose (BG) and TG lag time, and correlation with BG. A total of 4 animals were used. Eye irritation was monitored by Lissamine green dye and analyzed using image analysis software. Lag time was correlated with an oral glucose load while recording TG and BG readings. Correlation between TG and BG were plotted against one another to form a correlation diagram, using a Yellow Springs Instrument (YSI) and self-monitoring of blood glucose as the reference measurements. Finally, TG levels were calculated using analytically derived expressions. From repeated testing carried over the course of 12 months, little to no eye irritation was detected. TG fluctuations over time visually appeared to trace the same pattern as BG with an average lag times of 13 minutes. TG levels calculated from the device current measurements ranged from 4 to 20 mg/dL and correlated linearly with BG levels of 75-160 mg/dL (TG = 0.1723 BG = 7.9448 mg/dL; R 2 = .7544). The first steps were taken toward preliminary development of a sensor for self-monitoring of tear glucose (SMTG). No conjunctival irritation in any of the animals was noted. Lag time between TG and BG was found to be noticeable, but a quantitative modeling to correlate lag time in this study is unnecessary. Measured currents from the sensors and the calculated TG showed promising correlation to BG levels. Previous analytical bench marking showed BG and TG levels consistent with other literature. © 2014 Diabetes Technology Society.

On Preliminary Test Estimator for Median

OpenAIRE

Okazaki, Takeo; 岡崎, 威生

1990-01-01

The purpose of the present paper is to discuss about estimation of median with a preliminary test. Two procedures are presented, one uses Median test and the other uses Wilcoxon two-sample test for the preliminary test. Sections 3 and 4 give mathematical formulations of such properties, including mean square errors with one specified case. Section 5 discusses their optimal significance levels of the preliminary test and proposes their numerical values by Monte Carlo method. In addition to mea...

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 preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems

International Nuclear Information System (INIS)

Parlos, A.G.; Khan, E.U.; Frymire, R.; Negron, S.; Thomas, J.K.; Peddicord, K.L.

1991-01-01

Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the devleopment of highly compact space reactors. The calculations reported in this study include a neutronic design trade-off study using a two-dimensioinal neutron transport model, as well as a simplified one-dimensional thermal analysis of the reactor core. In arriving at the most desirable configuration, various options have been considered and analyzed, and their advantages/disadvantages have been compared. However, because of space limitation, only the most favorable reactor configuration is presented in this summary

Proposed preliminary definition of the disturbed-zone boundary appropriate for a repository at Yucca Mountain

International Nuclear Information System (INIS)

Langkopf, B.S.

1987-12-01

Some of the calculations that support the licensing of a repository for high-level radioactive waste will use the regulatory concept of a disturbed zone. The Nevada Nuclear Waste Storage Investigations (NNWSI) project must determine the location of the boundary of the disturbed zone for use in these calculations. This paper summarizes results of computer analyses and laboratory experiments and suggests a preliminary definition for the boundary of the disturbed zone for the unsaturated environment at Yucca Mountain. Although the intent of this paper is to define the boundary of the disturbed zone at the edge of significant changes in intrinsic hydrologic properties, there is no evidence of changes in intrinsic hydrologic properties that could significantly change the groundwater travel time from the repository to the water table. Such a result suggests that the disturbed zone at Yucca Mountain is of minimal extent. Because the analyses and experiments reviewed here indicate that there are a variety of changes near the waste package and because the results are subject to uncertainty, the preliminary suggestion for the extent of the disturbed zone is a value larger than the results themselves would suggest: the boundary is proposed to be a plane 10 m below the lower boundary of the waste packages. 88 refs., 12 figs., 5 tabs

Preliminary Modelling of Radiation Levels at the Fermilab PIP-II Linac

Energy Technology Data Exchange (ETDEWEB)

Lari, L. [CERN; Cerutti, F. [CERN; Esposito, L. S. [CERN; Baffes, C. [Fermilab; Dixon, S. J. [Fermilab; Mokhov, N. V. [Fermilab; Rakhno, I. [Fermilab; Tropin, I. S. [Fermilab

2018-04-01

PIP-II is the Fermilab's flagship project for providing powerful, high-intensity proton beams to the laboratory's experiments. The heart of PIP-II is an 800-MeV superconducting linac accelerator. It will be located in a new tunnel with new service buildings and connected to the present Booster through a new transfer line. To support the design of civil engineering and mechanical integration, this paper provides preliminary estimation of radiation level in the gallery at an operational beam loss limit of 0.1 W/m, by means of Monte Carlo calculations with FLUKA and MARS15 codes.

Model calculations as one means of satisfying the neutron cross-section requirements of the CTR program

International Nuclear Information System (INIS)

Gardner, D.G.

1975-01-01

A large amount of cross section and spectral information for neutron-induced reactions will be required for the CTR design program. To undertake to provide the required data through a purely experimental measurement program alone may not be the most efficient way of attacking the problem. It is suggested that a preliminary theoretical calculation be made of all relevant reactions on the dozen or so elements that now seem to comprise the inventory of possible construction materials to find out which are actually important, and over what energy ranges they are important. A number of computer codes for calculating cross sections for neutron induced reactions have been evaluated and extended. These will be described and examples will be given of various types of calculations of interest to the CTR program. (U.S.)

Manual for calculating critical loads of heavy metals for soils and surface waters; preliminary guidelines for environmental quality criteria, calculation methods and input data

NARCIS (Netherlands)

Vries, de W.; Bakker, D.J.

1996-01-01

Methodologies are described for calculating critical loads of lead, cadmium, copper, zinc, nickel, chromium and mercury for soils and surface waters. The aspects which are discussed are: selection of a computation model, determination of environmental-quality criteria for the metals, collection of

Preliminary Shielding Analysis for HCCB TBM Transport

Science.gov (United States)

Miao, Peng; Zhao, Fengchao; Cao, Qixiang; Zhang, Guoshu; Feng, Kaiming

2015-09-01

A preliminary shielding analysis on the transport of the Chinese helium cooled ceramic breeder test blanket module (HCCB TBM) from France back to China after being irradiated in ITER is presented in this contribution. Emphasis was placed on irradiation safety during transport. The dose rate calculated by MCNP/4C for the conceptual package design satisfies the relevant dose limits from IAEA that the dose rate 3 m away from the surface of the package containing low specific activity III materials should be less than 10 mSv/h. The change with location and the time evolution of dose rates after shutdown have also been studied. This will be helpful for devising the detailed transport plan of HCCB TBM back to China in the near future. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning

Energy Technology Data Exchange (ETDEWEB)

Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)

2014-05-15

The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.

Gas turbine designer computer program - a study of using a computer for preliminary design of gas turbines

Energy Technology Data Exchange (ETDEWEB)

Petersson, Rickard

1995-11-01

This thesis presents calculation schemes and theories for preliminary design of the fan, high pressure compressor and turbine of a gas turbine. The calculations are presented step by step, making it easier to implement in other applications. The calculation schemes have been implemented as a subroutine in a thermodynamic program. The combination of the thermodynamic cycle calculation and the design calculation turned out to give quite relevant results, when predicting the geometry and performance of an existing aero engine. The program developed is able to handle several different gas turbines, including those in which the flow is split (i.e. turbofan engines). The design process is limited to the fan, compressor and turbine of the gas turbine, the rest of the components have not been considered. Output from the program are main geometry, presented both numerically and as a scale plot, component efficiencies, stresses in critical points and a simple prediction of turbine blade temperatures. 11 refs, 21 figs, 1 tab

Preliminary evaluation of the expected radiation damage of the bayonet IFMIF back-plate

Energy Technology Data Exchange (ETDEWEB)

Frisoni, M. [Athena s.a.s, Via del Battiferro 3, I-40129 Bologna (Italy)], E-mail: manuela.frisoni@enea.it; Agostini, P. [ENEA CR Brasimone, Bacino del Brasimone 40032, Camugnano (Bolivia, Plurinational State of) (Italy); Fasanella, D. [Athena s.a.s, Via del Battiferro 3, I-40129 Bologna (Italy); Micciche, G. [ENEA CR Brasimone, Bacino del Brasimone 40032, Camugnano (Bolivia, Plurinational State of) (Italy)

2009-06-15

This paper summarises and discusses the results of a preliminary damage assessment of the non-seizure coating of the bayonet IFMIF back-plate. Neutron-induced kerma factors, dpa and gas production cross sections libraries were produced in a multigroup structure for neutron energies up to 60 MeV, by processing evaluated nuclear data files with NJOY-99.259 system. The material damage evaluations in terms of heat deposition, displacement and gas production rates were calculated using these libraries and compared with the values obtained using the data contained in the pointwise ACE format files of MCNP5 code package. The calculations were performed with MCNP5 code both using the McEnea and the McDelicious neutron source models to reproduce the energy-angle distributions of the neutrons produced in IFMIF d-Li interactions.

Purification, crystallization and preliminary X-ray diffraction studies of parakeet (Psittacula krameri) haemoglobin

International Nuclear Information System (INIS)

Jaimohan, S. M.; Naresh, M. D.; Arumugam, V.; Mandal, A. B.

2009-01-01

Parakeet (Psittacula krameri) haemoglobin has been purified and crystallized under low salt buffered conditions. Preliminary analysis of the crystal that belonged to monoclinic system (C2) is reported. Birds often show efficient oxygen management in order to meet the special demands of their metabolism. However, the structural studies of avian haemoglobins (Hbs) are inadequate for complete understanding of the mechanism involved. Towards this end, purification, crystallization and preliminary X-ray diffraction studies have been carried out for parakeet Hb. Parakeet Hb was crystallized as the met form in low-salt buffered conditions after extracting haemoglobin from crude blood by microcentrifugation and purifying the sample by column chromatography. Good-quality crystals were grown from 10% PEG 3350 and a crystal diffracted to about 2.8 Å resolution. Preliminary diffraction data showed that the Hb crystal belonged to the monoclinic system (space group C2), with unit-cell parameters a = 110.68, b = 64.27, c = 56.40 Å, β = 109.35°. Matthews volume analysis indicated that the crystals contained a half-tetramer in the asymmetric unit

Methodology and a preliminary data base for examining the health risks of electricity generation from uranium and coal fuels

Energy Technology Data Exchange (ETDEWEB)

El-Bassioni, A.A.

1980-08-01

An analytical model was developed to assess and examine the health effects associated with the production of electricity from uranium and coal fuels. The model is based on a systematic methodology that is both simple and easy to check, and provides details about the various components of health risk. A preliminary set of data that is needed to calculate the health risks was gathered, normalized to the model facilities, and presented in a concise manner. Additional data will become available as a result of other evaluations of both fuel cycles, and they should be included in the data base. An iterative approach involving only a few steps is recommended for validating the model. After each validation step, the model is improved in the areas where new information or increased interest justifies such upgrading. Sensitivity analysis is proposed as the best method of using the model to its full potential. Detailed quantification of the risks associated with the two fuel cycles is not presented in this report. The evaluation of risks from producing electricity by these two methods can be completed only after several steps that address difficult social and technical questions. Preliminary quantitative assessment showed that several factors not considered in detail in previous studies are potentially important. 255 refs., 21 figs., 179 tabs.

How Accurately Can We Calculate Neutrons Slowing Down In Water ?

International Nuclear Information System (INIS)

Cullen, D E; Blomquist, R; Greene, M; Lent, E; MacFarlane, R; McKinley, S; Plechaty, E; Sublet, J C

2006-01-01

We have compared the results produced by a variety of currently available Monte Carlo neutron transport codes for the relatively simple problem of a fast source of neutrons slowing down and thermalizing in water. Initial comparisons showed rather large differences in the calculated flux; up to 80% differences. By working together we iterated to improve the results by: (1) insuring that all codes were using the same data, (2) improving the models used by the codes, and (3) correcting errors in the codes; no code is perfect. Even after a number of iterations we still found differences, demonstrating that our Monte Carlo and supporting codes are far from perfect; in particularly we found that the often overlooked nuclear data processing codes can be the weakest link in our systems of codes. The results presented here represent the today's state-of-the-art, in the sense that all of the Monte Carlo codes are modern, widely available and used codes. They all use the most up-to-date nuclear data, and the results are very recent, weeks or at most a few months old; these are the results that current users of these codes should expect to obtain from them. As such, the accuracy and limitations of the codes presented here should serve as guidelines to code users in interpreting their results for similar problems. We avoid crystal ball gazing, in the sense that we limit the scope of this report to what is available to code users today, and we avoid predicting future improvements that may or may not actual come to pass. An exception that we make is in presenting results for an improved thermal scattering model currently being testing using advanced versions of NJOY and MCNP that are not currently available to users, but are planned for release in the not too distant future. The other exception is to show comparisons between experimentally measured water cross sections and preliminary ENDF/B-VII thermal scattering law, S(α,β) data; although these data are strictly preliminary

Preliminary study for unified management of CANDU safety codes and construction of database system

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae

2003-03-01

It is needed to develop the Graphical User Interface(GUI) for the unified management of CANDU safety codes and to construct database system for the validation of safety codes, for which the preliminary study is done in the first stage of the present work. The input and output structures and data flow of CATHENA and PRESCON2 are investigated and the interaction of the variables between CATHENA and PRESCON2 are identified. Furthermore, PC versions of CATHENA and PRESCON2 codes are developed for the interaction of these codes and GUI(Graphic User Interface). The PC versions are assessed by comparing the calculation results with those by HP workstation or from FSAR(Final Safety Analysis Report). Preliminary study on the GUI for the safety codes in the unified management system are done. The sample of GUI programming is demonstrated preliminarily. Visual C++ is selected as the programming language for the development of GUI system. The data for Wolsong plants, reactor core, and thermal-hydraulic experiments executed in the inside and outside of the country, are collected and classified following the structure of the database system, of which two types are considered for the final web-based database system. The preliminary GUI programming for database system is demonstrated, which is updated in the future work

Preliminary Ecotoxicity and Biodegradability Assessment of Metalworking Fluids

Science.gov (United States)

Gerulová, Kristína; Amcha, Peter; Filická, Slávka

2010-01-01

The main aim of this study was to evaluate the potential of activated sludge from sewage treatment plant to degrade selected MWFs (ecotoxicity to bacterial consortium) and to evaluate the ecotoxicity by Lemna minor-higher plant. After evaluating the ecotoxicity, biodegradations rate with activated sludge was assessed on the basis of COD measurement. Preliminary study of measuring the ecotoxicity according to OECD 221 by Lemna minor shows effective concentration of Emulzin H at the rate of 81.6 mg l-1, for Ecocool 82.9 mg l-1, for BC 25 about 99.3 mg l-1, and for Dasnobor about 97.3 mg l-1. Preliminary study of measuring the ecotoxicity by bacterial consortium according to OECD 209 (STN EN ISO 8192) shows effective concentration of Blasocut BC 25 at the rate 227.4 mg l-1. According to OECD 302B, the biodegradations level of Emulzin H, Ecocool and BC 25 achieved 80% in 10 days. It can be stated that these MWFs have potential to ultimate degradation, but the statement has to be confirmed by a biodegradability test with other parameters than COD, which exhibits some disadvantages in testing O/W emulsions.

LM-OSL from single grains of quartz: A preliminary study

DEFF Research Database (Denmark)

Bulur, E.; Duller, G.A.T.; Solongo, S.

2002-01-01

the easy-to-bleach component, those with only the hard-to-bleach component, and those exhibiting all components. The results of this preliminary study show that LM-OSL experiments carried out at the single grain level may give important insights into the luminescence properties observed when viewing...

Preliminary results of statistical dynamic experiments on a heat exchanger

International Nuclear Information System (INIS)

Corran, E.R.; Cummins, J.D.

1962-10-01

The inherent noise signals present in a heat exchanger have been recorded and analysed in order to determine some of the statistical dynamic characteristics of the heat exchanger. These preliminary results show that the primary side temperature frequency response may be determined by analysing the inherent noise. The secondary side temperature frequency response and cross coupled temperature frequency responses between primary and secondary are poorly determined because of the presence of a non-stationary noise source in the secondary circuit of this heat exchanger. This may be overcome by correlating the dependent variables with an externally applied noise signal. Some preliminary experiments with an externally applied random telegraph type of signal are reported. (author)

Calculation of aerodynamics of aerosol filter designs for cleaning of heavy liquid metal cooler reactor gas loops

International Nuclear Information System (INIS)

Valery P Melnikov; Pyotr N Martynov; Albert K Papovyants; Ivan V Yagodkin

2005-01-01

Full text of publication follows: One of the basic performances of aerosol filters is the aerodynamic resistance to the flow of gaseous medium to be cleaned. Calculation of the aerodynamics of aerosol filters in reference to the gas loops of reactor installations with heavy liquid metal coolant (HLMC) allows the design of the structural components of filters to be optimized to provide minimum initial resistance values. It is established that owing to various factors aerosol particles of different concentration and disperse composition are present always in the gas spaces of heavy liquid metal cooled reactor gas loops. To prevent the negative effect of aerosols on the equipment of the gas loops, it is reasonable to use filters of multistep design with sections of preliminary and fine cleaning to catch micron and submicron particles, respectively. A computer program and technique have been developed to evaluate the aerodynamics of folded aerosol filters for different parameters of their structural components, taking account of the aerosol spectrum and concentration. The algorithm of the calculation is presented by the example of a two-step design assembled in single vessel; the filter dimensions and pattern of the air flow to be cleaned are determined under the given boundary conditions. The evaluation of the aerodynamic resistance of filters was performed with consideration for local resistances and resistances of all the structural components of the filter (sudden constriction, expansion, the flow in air channels, filtering material and so on). Correlations have been derived for the resistance of air channels, filtering materials of preliminary and fine cleaning sections as a function of such parameters as the section depth (50-500 mm), the height of separators (3,5-20 mm), the filtering surface area (1,5-30 m 2 ). Based on the calculation results, the auto-similarity domain was brought out for the minimal values of filter resistances as a function of the ratio of

Magma addition rates in continental arcs: New methods of calculation and global implications

Science.gov (United States)

Ratschbacher, B. C.; Paterson, S. R.

2017-12-01

The transport of mass, heat and geochemical constituents (elements and volatiles) from the mantle to the atmosphere occurs via magma addition to the lithosphere. Calculation of magma addition rates (MARs) in continental arcs based on exposed proportions of igneous arc rocks is complex and rarely consistently determined. Multiple factors influence MAR calculations such as crust versus mantle contributions to magmas, a change in MARs across the arc and with depths throughout the arc crustal column, `arc tempos' with periods of high and low magmatic activity, the loss of previous emplaced arc rocks by subsequent magmatism and return to the mantle, arc migration, variations in the intrusive versus extrusive additions and evolving arc widths and thicknesses during tectonism. All of these factors need to be considered when calculating MARs.This study makes a new attempt to calculate MARs in continental arcs by studying three arc sections: the Famatinian arc, Argentina, the Sierra Nevada batholith, California and the Coast Mountain batholith, Washington and British Columbia. Arcs are divided into fore-arc, main arc and back arc sections and `boxes' with a defined width, length and thickness spanning upper middle and lower crustal levels are assigned to each section. Representative exposed crustal slices for each depth are then used to calculate MARs based on outcrop proportions for each box. Geochemical data is used to infer crustal recycling percentages and total thickness of the arc. Preliminary results show a correlation between MARs, crustal thicknesses and magmatic flare-up durations. For instance, the Famatinian arc shows a strong decrease in MARs between the main arc section (9.4 km3/Ma/arc-km) and the fore-arc (0.61 km3/Ma/arc-km) and back-arc (1.52 km3/Ma/arc-km) regions and an increase in the amount of magmatism with depth.Global MARs over geologic timescales have the potential to investigate mantle melt generation rates and the volatile outgassing contribution

Preliminary test results for post irradiation examination on the HTTR fuel

International Nuclear Information System (INIS)

Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

2007-01-01

The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

Preliminary Analysis on Linac Oscillation Data LI05-19 and Wake Field Energy Loss in FACET Commissioning 2012

Energy Technology Data Exchange (ETDEWEB)

Sun, Yipeng; /SLAC

2012-07-23

In this note, preliminary analysis on linac ocsillation data in FACET linac LI05-09 plus LI11-19 is presented. Several quadrupoles are identified to possibly have different strength, compared with their designed strength in the MAD optics model. The beam energy loss due to longitudinal wake fields in the S-band linac is also analytically calculated, also by LITRACK numerical simulations.

CONTAIN calculations; CONTAIN-Rechnungen

Energy Technology Data Exchange (ETDEWEB)

Scholtyssek, W.

1995-08-01

In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident `medium-sized leak in the cold leg`, especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

Preliminary heat flow map of Europe. Explanatory text

Energy Technology Data Exchange (ETDEWEB)

Cermak, V.; Hurtig, E.

1977-08-08

A preliminary heat flow map of Europe was prepared, based on data contained in 401 references. The map was prepared on a scale of 1:5,000,000 and shows broad-scale geological structure (e.g., platforms, shields, foredeeps) and specialized rock suites (ophiolites, volcanites). Primary faults and thrust faults are indicated, and contours showing the depth of crystalline basement are given. Heat flow is plotted using 10.0 mW/m/sup 2/ isotherms. The accompanying explanatory text describes data acquisition and techniques of correction, and discusses some implications of the results.

Detailed characterization and preliminary adsorption model for materials for an intermediate-scale reactive-transport experiment

International Nuclear Information System (INIS)

Ward, D.B.; Bryan, C.R.

1994-01-01

An experiment involving migration of fluid and tracers (Li, Br, Ni) through a 6-m-high x 3-m-dia caisson Wedron 510 sand, is being carried out for Yucca Mountain Site Characterization Project. Sand's surface chemistry of the sand was studied and a preliminary surface-complexation model of Ni adsorption formulated for transport calculations. XPS and leaching suggest that surface of the quartz sand is partially covered by thin layers of Fe-oxyhydroxide and Ca-Mg carbonate and by flakes of kaolinite. Ni adsorption by the sand is strongly pH-dependent, showing no adsorption at pH 5 and near-total adsorption at pH 7. Location of adsorption edge is independent of ionic strength and dissolved Ni concentration; it is shifted to slightly lower pH with higher pCO2 and to slightly higher pH by competition with Li. Diminished adsorption at alkiline pH with higher pCO2 implies formation of dissolved Ni-carbonato complexes. Ni adsorption edges for goethite and quartz, two components of the sand were also measured. Ni adsorption on pure quartz is only moderately pH-dependent and differs in shape and location from that of the sand, whereas Ni adsorption by goethite is strongly pH-dependent. A triple-layer surface-complexation model developed for goethite provides a good fit to the Ni-adsorption curve of the sand. Based on this model, the apparent surface area of the Fe-oxyhydroxide coating is estimated to be 560 m 2 /g, compatible with its occurrence as amorphous Fe-oxyhydroxide. Potentiometric titrations on sand also differ from pure quartz and suggest that effective surface area of sand may be much greater than that measured by N 2 -BET gas adsorption. Attempts to model the adsorption of bulk sand in terms of properties of pure end member components suggest that much of the sand surface is inert. Although the exact Ni adsorption mechanisms remain ambiguous, this preliminary adsorption model provides an initial set of parameters that can be used in transport calculations

Hellenic Amateur Astronomy Association's activities: Preliminary results on Perseids 2010

Science.gov (United States)

Maravelias, G.

2011-01-01

Preliminary results on the Perseids 2010 are presented. Visual and video observations were obtained by the author and a first reduction of the visual data shows that a maximum of ZHR ~120 was reached during the night 12-13 of August 2010. Moreover, a video setup was tested (DMK camera and UFO Capture v2) and the results show that, under some limitations, valuable data can be obtained.

23 CFR 645.109 - Preliminary engineering.

Science.gov (United States)

2010-04-01

... 23 Highways 1 2010-04-01 2010-04-01 false Preliminary engineering. 645.109 Section 645.109 Highways FEDERAL HIGHWAY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION ENGINEERING AND TRAFFIC OPERATIONS UTILITIES Utility Relocations, Adjustments, and Reimbursement § 645.109 Preliminary engineering. (a) As...

Preliminary design study of a steady state tokamak device

International Nuclear Information System (INIS)

Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

1992-09-01

Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

International Nuclear Information System (INIS)

1990-02-01

The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics

International Nuclear Information System (INIS)

Seker, V.; Thomas, J.W.; Downar, T.J.

2007-01-01

A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR'. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. Preliminary testing of the code was performed using a 3x3 PWR pin-cell problem. The preliminary results are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code DeCART. Good agreement in the k eff and the power profile was observed. Increased computational capabilities and improvements in computational methods have accelerated interest in high fidelity modeling of nuclear reactor cores during the last several years. High-fidelity has been achieved by utilizing full core neutron transport solutions for the neutronics calculation and computational fluid dynamics solutions for the thermal-hydraulics calculation. Previous researchers have reported the coupling of 3D deterministic neutron transport method to CFD and their application to practical reactor analysis problems. One of the principal motivations of the work here was to utilize Monte Carlo methods to validate the coupled deterministic neutron transport

Preliminary design county plan Zeeland

International Nuclear Information System (INIS)

1987-01-01

The preliminary design 'Streekplan Zeeland' (Country plan Zeeland, with regard to the location of additional nuclear power plants in Zeeland, the Netherlands) has passed through a consultation and participation round. Thereupon 132 reactions have been received. These have been incorporated and answered in two notes. This proposal deals with the principal points of the preliminary design and treats also the remarks of the committees Environmental (town and country) Planning (RO), Provincial (town and country) Planning Committee (PPC) and Association of Communities of Zeeland (VZG), on the reply notes. The preliminary design with the modifications, collected in appendix 3, is proposed to be the starting point in the drawing-up of the design-country-plan. This design subsequently will pass the formal country-plan procedure. (author). 1 fig

Selecting and calculating joint operation of oil and petroleum gas collection systems, and mechanized production methods

Energy Technology Data Exchange (ETDEWEB)

Guseva, L S; D' yachuk, A I; Davydova, L V; Maslov, V P; Salyautdinova, R M; Suslov, V M

1979-01-01

The possibility is examined of formalizing the indicated procedure in the process of performing step by step calculations. At the first step, considering limitations imposed by the dominating parameters, preliminary selection is performed of the acceptable combination of the type of collection system and methods of mechanized production for the development conditions examined. The second step provides for physical simulation at a well of an experimental section of time-variable conditions of field development. The values of the technological indices thus defined are then considered to be reliable information for technico-economic calculations. Parallel research is done on the technological features of operation of the collection systems chosen and their individual elements (pipeline system, separation units, etc.), which the experimental section is fitted with beforehand. Material is given which illustrates in detail the basic assumptions of the technique proposed and the calculation procedure.

Calculating Cluster Masses via the Sunyaev-Zel'dovich Effect

Science.gov (United States)

Lindley, Ashley; Landry, D.; Bonamente, M.; Joy, M.; Bulbul, E.; Carlstrom, J. E.; Culverhouse, T. L.; Gralla, M.; Greer, C.; Hawkins, D.; Lamb, J. W.; Leitch, E. M.; Marrone, D. P.; Miller, A.; Mroczkowski, T.; Muchovej, S.; Plagge, T.; Woody, D.

2012-05-01

Accurate measurements of the total mass of galaxy clusters are key for measuring the cluster mass function and therefore investigating the evolution of the universe. We apply two new methods to measure cluster masses for five galaxy clusters contained within the Brightest Cluster Sample (BCS), an X-ray luminous statistically complete sample of 35 clusters at z=0.15-0.30. These methods distinctively use only observations of the Sunyaev-Zel'dovich (SZ) effect, for which the brightness is redshift independent. At the low redshifts of the BCS, X-ray observations can easily be used to determine cluster masses, providing convenient calibrators for our SZ mass calculations. These clusters have been observed with the Sunyaev-Zel'dovich Array (SZA), an interferometer that is part of the Combined Array for Research in Millimeter-wave Astronomy (CARMA) that has been optimized for accurate measurement of the SZ effect in clusters of galaxies at 30 GHz. One method implements a scaling relation that relates the integrated pressure, Y, as determined by the SZ observations to the mass of the cluster calculated via optical weak lensing. The second method makes use of the Virial theorem to determine the mass given the integrated pressure of the cluster. We find that masses calculated utilizing these methods within a radius r500 are consistent with X-ray masses, calculated by manipulating the surface brightness and temperature data within the same radius, thus concluding that these are viable methods for the determination of cluster masses via the SZ effect. We present preliminary results of our analysis for five galaxy clusters.

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

A FIRST APPROXIMATION CALCULATION OF AIR CUSHION CHASSIS WEIGHT OF TRANSPORT AIRPLANE

Directory of Open Access Journals (Sweden)

2016-01-01

Full Text Available This article describes a first approximation of a weighted estimate of air cushion chassis. The algorithm for calculating the weight of air cushion chassis allows not only to estimate the mass of the chassis to a first approximation, but also to conduct a preliminary analysis of the influence of various parameters of the aircraft and the chassis on the weight of the aircraft at the stage of before designing. The algorithm can be expanded to include additional design decisions, such as the transformation of the fuselage, increasing the air cushion chassis canopy due to extensions, center of gravity, etc.

Preliminary shielding calculation for the system of CyberKnife robotic radiosurgery; Calculo de blindagem preliminar para o sistema de radiocirurgia robotica CyberKnife

Energy Technology Data Exchange (ETDEWEB)

Toreti, Dalila; Xavier, Clarice; Moura, Fabio, E-mail: clarice.xavier@rem.ind.b, E-mail: fabio.moura@rem.ind.b [REM Industria e Comercio Ltda., Sao Paulo, SP (Brazil)

2011-10-26

The CyberKnife robotic system uses a manipulator with six grade of freedom for positioning a 6 MV Linac accelerator for treatment of lesions. This paper presents calculations for a standard room, with 200 cm of thickness walls primary, build for a CyberKnife system, and calculations for a room originally designed for a Linac conventional (with gantry), with secondary barriers of 107 cm thickness. After the realization of shielding for both rooms, the results shown that walls of standard room with 200 cm thickness are adequate for the secondary shield, and for a room with a conventional Linac, from all six evaluated points, two would require additional shielding of nine cm and four cm of concrete with 2.4 g/cubic cm. This shows that the CyberKnife system can be installed in a originally designed room for a conventional Linac with neither restrict nor any shielding, since no incidence of beams on the secondary barriers is existent

Exploratory shaft facility preliminary designs - Paradox Basin. Technical report

International Nuclear Information System (INIS)

1983-09-01

The purpose of the Preliminary Design Report, Paradox Basin, is to provide a description of the preliminary design for an Exploratory Shaft Facility in the Paradox Basin, Utah. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling Method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers is included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description, and Construction Cost Estimate. 30 references

Design and preliminary results of a fuel flexible industrial gas turbine combustor

Science.gov (United States)

Novick, A. S.; Troth, D. L.; Yacobucci, H. G.

1981-01-01

The design characteristics are presented of a fuel tolerant variable geometry staged air combustor using regenerative/convective cooling. The rich/quench/lean variable geometry combustor is designed to achieve low NO(x) emission from fuels containing fuel bound nitrogen. The physical size of the combustor was calculated for a can-annular combustion system with associated operating conditions for the Allison 570-K engine. Preliminary test results indicate that the concept has the potential to meet emission requirements at maximum continuous power operation. However, airflow sealing and improved fuel/air mixing are necessary to meet Department of Energy program goals.

Preliminary design of the beam transport system for the Milan biomedical cyclotron

International Nuclear Information System (INIS)

Silari, M.

1988-01-01

This report illustrates the preliminary design of the beam transport system for the Scanditronix MC40 cyclotron to be installed in Milan. The Cyclotron will be dedicated to biomedical research and the different experimental conditions that could occur will require a beam transport system flexible enough so as to deliver beams with the specified characteristics. The report describes the computer codes used, the calculations performed and the results obtained. The complete configuration of the beam lines serving the first two target rooms is given, together with typical beam profiles and the emittance ellipse variation along the transfer channels

PRELIMINARY PHYTOCHEMICAL INVESTIGATION AND THIN LAYER CHROMATOGRAPHY OF RHEUM EMODI

OpenAIRE

Mir Ashfaq Ahmad; K. W. Shah; Showkat Ahmad Wani

2012-01-01

Preliminary phytochemical investigation of aqueous and methanolic rhizome extracts of Rheum emodi followed by their TLC profiling were carried out. Phytochemical analysis reveals the presence of diverse groups of phytoconstituents in two different extracts (aqueous and methanolic rhizome extracts). Chemical constituents also show different Rf values in two different solvent systems.

Infinite slab-shield dose calculations

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

Present status of the physical-neutronic calculation system for nuclear fuel management at the Juragua Power Plant WWER-440 reactors

International Nuclear Information System (INIS)

Garcia H, Carlos R.; Milian, Daniel

1997-01-01

In this paper the most remarkable results obtained in Cuba, for the Juragua Nuclear Power Plant (JNPP) in core fuel management, are shown. The main characteristic of the codes used to solve the usual sequence of neutron-physical calculations available in our code library are reported. The codes validation was based on estimative of their accuracy by inter comparing calculated and WWER-440 NPP's operation data. A brief summary of the extensive calculations carried out for the JNPP Preliminary Safety Report elaboration is presented. This report is one of the requisites demanded by the Cuban Nuclear Agency for issuing the General Permission for restart works in the plant. (author). 12 refs., 4 figs., 6 tabs

Two-dimensional computer simulation of hypervelocity impact cratering: some preliminary results for Meteor Crater, Arizona

Energy Technology Data Exchange (ETDEWEB)

Bryan, J.B.; Burton, D.E.; Cunningham, M.E.; Lettis, L.A. Jr.

1978-06-01

A computational approach used for subsurface explosion cratering was extended to hypervelocity impact cratering. Meteor (Barringer) Crater, Arizona, was selected for the first computer simulation because it is one of the most thoroughly studied craters. It is also an excellent example of a simple, bowl-shaped crater and is one of the youngest terrestrial impact craters. Initial conditions for this calculation included a meteorite impact velocity of 15 km/s, meteorite mass of 1.67 x 10/sup 8/ kg, with a corresponding kinetic energy of 1.88 x 10/sup 16/ J (4.5 megatons). A two-dimensional Eulerian finite difference code called SOIL was used for this simulation of a cylindrical iron projectile impacting at normal incidence into a limestone target. For this initial calculation, a Tillotson equation-of-state description for iron and limestone was used with no shear strength. Results obtained for this preliminary calculation of the formation of Meteor Crater are in good agreement with field measurements. A color movie based on this calculation was produced using computer-generated graphics. 19 figures, 5 tables, 63 references.

Two-dimensional computer simulation of hypervelocity impact cratering: some preliminary results for Meteor Crater, Arizona

International Nuclear Information System (INIS)

Bryan, J.B.; Burton, D.E.; Cunningham, M.E.; Lettis, L.A. Jr.

1978-06-01

A computational approach used for subsurface explosion cratering was extended to hypervelocity impact cratering. Meteor (Barringer) Crater, Arizona, was selected for the first computer simulation because it is one of the most thoroughly studied craters. It is also an excellent example of a simple, bowl-shaped crater and is one of the youngest terrestrial impact craters. Initial conditions for this calculation included a meteorite impact velocity of 15 km/s, meteorite mass of 1.67 x 10 8 kg, with a corresponding kinetic energy of 1.88 x 10 16 J (4.5 megatons). A two-dimensional Eulerian finite difference code called SOIL was used for this simulation of a cylindrical iron projectile impacting at normal incidence into a limestone target. For this initial calculation, a Tillotson equation-of-state description for iron and limestone was used with no shear strength. Results obtained for this preliminary calculation of the formation of Meteor Crater are in good agreement with field measurements. A color movie based on this calculation was produced using computer-generated graphics. 19 figures, 5 tables, 63 references

XeCl Excimer Laser For Micro - Machining Of Materials: Preliminary Theoretical And Experimental Works.

Science.gov (United States)

Iwanejko, Leszek; Pokora, Ludwik; Stefanski, Miroslaw; Ujda, Zbigniew

1987-10-01

The paper presents the results of preliminary investigations, both theoretical and experimental, of XeC1 excimer laser pumped by transverse electric discharge with UU preionization. The medium was a mixture of gases He-Xe-HC1. A theoretical model of the XeC1 laser was worked out and a lot of laser parameters calculations were done. In the same time an excimer laser operating on the mixture He-Xe-HC1 was started, the generation of laser radiation was of energy about 20mJ.

The Preliminary GAMMA Code Thermal hydraulic Analysis for the Steady State of HTR-10 Initial Core

Energy Technology Data Exchange (ETDEWEB)

Jun, Ji Su; Lim, Hong Sik; Lee, Won Jae

2006-07-15

This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a

Adaptation of GRS calculation codes for Soviet reactors

International Nuclear Information System (INIS)

Langenbuch, S.; Petri, A.; Steinborn, J.; Stenbok, I.A.; Suslow, A.I.

1994-01-01

The use of ATHLET for incident calculation of WWER has been tested and verified in numerous calculations. Further adaptation may be needed for the WWER 1000 plants. Coupling ATHLET with the 3D nuclear model BIPR-8 for WWER cores clearly improves studies of the influence of neutron kinetics. In the case of FBMK reactors ATHLET calculations show that typical incidents in the complex RMBK reactors can be calculated even though verification still has to be worked on. Results of the 3D-core model QUABOX/CUBBOX-HYCA show good correlation of calculated and measured values in reactor plants. Calculations carried out to date were used to check essential parameters influencing RBMK core behaviour especially dependence of effective voidre activity on the number of control rods. (orig./HP) [de

32 CFR 644.30 - Preliminary real estate work.

Science.gov (United States)

2010-07-01

... 32 National Defense 4 2010-07-01 2010-07-01 true Preliminary real estate work. 644.30 Section 644... PROPERTY REAL ESTATE HANDBOOK Project Planning Military (army and Air Force) and Other Federal Agencies § 644.30 Preliminary real estate work. (a) Preliminary real estate work is defined as that action taken...

Monitoring of German Fertility: Estimation of Monthly and Yearly Total Fertility Rates on the Basis of Preliminary Monthly Data

Directory of Open Access Journals (Sweden)

Gabriele Doblhammer

2011-02-01

Full Text Available This paper introduces a set of methods for estimating fertility indicators in the absence of recent and short-term birth statistics. For Germany, we propose a set of straightforward methods that allow for the computation of monthly and yearly total fertility rates (mTFR on the basis of preliminary monthly data, including a confidence interval. The method for estimating most current fertility rates can be applied when no information on the age structure and the number of women exposed to childbearing is available. The methods introduced in this study are useful for calculating monthly birth indicators, with minimal requirements for data quality and statistical effort. In addition, we suggest an approach for projecting the yearly TFR based on preliminary monthly information up to June.

Preliminary safety criteria for organic watch list tanks at the Hanford site

International Nuclear Information System (INIS)

Webb, A.B.; Stewart, J.L.; Turner, O.A.; Plys, M.G.; Malinovic, B.; Grigsby, J.M.; Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J.

1995-11-01

Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended

Preliminary safety criteria for organic watch list tanks at the Hanford site

Energy Technology Data Exchange (ETDEWEB)

Webb, A.B.; Stewart, J.L.; Turner, O.A. [Westinghouse Hanford Co., Richland, WA (United States); Plys, M.G.; Malinovic, B. [Fauske and Associates, Inc., Burr Ridge, IL (United States); Grigsby, J.M. [G & P Consulting, Inc. (United States); Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J. [Pacific Northwest Lab., Portland, OR (United States)

1995-11-01

Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended.

Coupled-cluster calculations for ground and excited states of closed- and open-shell nuclei using methods of quantum chemistry

International Nuclear Information System (INIS)

Wloch, Marta; Gour, Jeffrey R; Piecuch, Piotr; Dean, David J; Hjorth-Jensen, Morten; Papenbrock, Thomas

2005-01-01

We discuss large-scale ab initio calculations of ground and excited states of 16 O and preliminary calculations for 15 O and 17 O using coupled-cluster methods and algorithms developed in quantum chemistry. By using realistic two-body interactions and the renormalized form of the Hamiltonian obtained with a no-core G-matrix approach, we are able to obtain the virtually converged results for 16 O and promising results for 15 O and 17 O at the level of two-body interactions. The calculated properties other than binding and excitation energies include charge radius and charge form factor. The relatively low costs of coupled-cluster calculations, which are characterized by the low-order polynomial scaling with the system size, enable us to probe large model spaces with up to seven or eight major oscillator shells, for which nontruncated shell-model calculations for nuclei with A = 15-17 active particles are presently not possible

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.

The dynamic capacity calculation method and the flood control ability of the Three Gorges Reservoir

Science.gov (United States)

Zhang, Shanghong; Jing, Zhu; Yi, Yujun; Wu, Yu; Zhao, Yong

2017-12-01

To evaluate the flood control ability of a river-type reservoir, an accurate simulation method for the flood storage, discharge process, and dynamic capacity of the reservoir is important. As the world's largest reservoir, the storage capacity and flood control capacity of the Three Gorges Reservoir (TGR) has attracted widespread interest and academic debate for nearly 20 years. In this study, a model for calculating the dynamic capacity of a river-type reservoir is established based on data from 394 river cross sections and 2.5-m resolution digital elevation model (DEM) data of the TGR area. The storage capacity and flood control capacity of the TGR were analysed based on the scheduling procedures of a normal impoundment period. The results show that the static capacity of the TGR is 43.43 billion m3, the dynamic flood control capacity is 22.45 billion m3, and the maximum floodwater flow regulated by the dynamic capacity at Zhicheng is no more than 67,700 m3/s. This study supply new simulation method and up-to-date high-precision data to discuss the 20 years debate, and the results reveal the TGR design is conservative for flood control according to the Preliminary Design Report of the Three Gorges Project. The dynamic capacity calculation method used here can provide a reference for flood regulation of large river-type reservoirs.

Plutonium Immobilization Can Loading Preliminary Specifications

Energy Technology Data Exchange (ETDEWEB)

Kriikku, E.

1998-11-25

This report discusses the Plutonium Immobilization can loading preliminary equipment specifications and includes a process block diagram, process description, equipment list, preliminary equipment specifications, plan and elevation sketches, and some commercial catalogs. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.

MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

Energy Technology Data Exchange (ETDEWEB)

Petoukhov, S.M.; Mathew, P.M

2001-10-01

To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

Preliminary Results Of LOCA Problem For APR1400 Reactor

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Le Thi Thu; Vo Thi Huong; Le Van Hong

2011-01-01

Several features of NPP with APR1400 nuclear reactor during a loss of coolant accident (LOCA) are investigated in this study. The report describes some main design characteristics of an engineering safety systems of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Large Break LOCA accident has been analyzed and evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with broken area of 0.0465 m 2 in case of high pressure safety injection system (HPSI) failed to operate or 2 and 4 HPSI pumps are activated. The preliminary results of this work is a part of collaboration between INST researchers and KAERI experts in using RELAP tool for safety analysis of NPPs. (author)

The effects of electric forces on dust lifting: Preliminary studies with a numerical model

International Nuclear Information System (INIS)

Kok, J F; Renno, N O

2008-01-01

Atmospheric dust aerosols affect the Earth's climate by scattering and absorbing radiation and by modifying cloud properties. Recent experiments have indicated that electric fields produced in dusty phenomena such as dust storms and dust devils could enhance the emission of dust aerosols. However, the generation of electric fields in dusty phenomena is poorly understood. To address this problem, we present results from the first physically-based numerical model of electric fields in dust lifting. Our model calculates the motion and collisions of air-borne particles, as well as the charge transfer during these collisions. This allows us to simulate the formation of electric fields as a function of physical parameters, such as wind stress and soil properties. Preliminary model results show that electric fields can indeed enhance the lifting of soil particles. Moreover, they suggest that strong electric fields could trigger a positive feedback because increases in the concentration of charged particles strengthen the original electric field, which in turn lifts additional surface particles. We plan to further test and calibrate our model with experimental data.

Development of a nuclear spallation simulation code and calculations of primary spallation products

International Nuclear Information System (INIS)

Nishida, Takahiko; Nakahara, Yasuaki; Tsutsui, Tsuneo

1986-08-01

In order to make evaluations of computational models for the nuclear spallation reaction from a nuclear physics point of view, a simulation code NUCLEUS has been developed by modifying and combining the Monte Carlo codes NMTC/JAERI and NMTA/JAERI for calculating only the nuclear spallation reaction (intranuclear cascade + evaporation and/or fast fission) between a nucleus and a projectile without taking into consideration of internuclear transport. New several plotting routines have been provided for the rapid process of much more event data, obtained by using the ARGUS plotting system. The results obtained by our code can be directly compared with the experimental results using by thin foil experiments in which internuclear multiple collisions have little effects, and will serve to upgrade the calculational methods and the values of nuclear parameters currently used in the calculations. Some discussions are done about the preliminary computational results obtained by using NUCLEUS. The mass distribution and charge dispersion of reaction products are examined in some detail for the nuclear spallation reaction between incident protons and target nuclei, such as U, Pb and Ag, in the energy range from 0.5 GeV to 3.0 GeV. These results show that the distribution of reaction products ceases to change its form as the proton energy increases over about 2 GeV. The same tendency is seen in the energy dependence of the number of primary particles emitted from a nucleus. After spallation reactions, a variety of nuclei, especially many neutron deficient nuclides with nuclear charges nearly equal to ones of a target nucleus, are produced. Due to their short lifetime most of them will change to stable nuclides in due time. Finally, some important issues are discussed to improve the present simulation method. (author)

Online Distance Education at the Universiti Sains Malaysia, Malaysia: Preliminary Perceptions.

Science.gov (United States)

Idrus, Rozhan Mohammed; Lateh, Habibah Hj

2000-01-01

Presents the instructional design aspects of a multimedia course delivered online via the Internet in a distance learning program at the Universiti Sains Malaysia. Reports results of a preliminary study that showed student perceptions toward online distance education focused on accessibility and presentation rather than pedagogical techniques and…

Combining scenarios in a calculation of the overall probability distribution of cumulative releases of radioactivity from the Waste Isolation Pilot Plant, southeastern New Mexico

International Nuclear Information System (INIS)

Tierney, M.S.

1991-11-01

The Waste Isolation Pilot Plant (WIPP), in southeastern New Mexico, is a research and development facility to demonstrate safe disposal of defense-generated transuranic waste. The US Department of Energy will designate WIPP as a disposal facility if it meets the US Environmental Protection Agency's standard for disposal of such waste; the standard includes a requirement that estimates of cumulative releases of radioactivity to the accessible environment be incorporated in an overall probability distribution. The WIPP Project has chosen an approach to calculation of an overall probability distribution that employs the concept of scenarios for release and transport of radioactivity to the accessible environment. This report reviews the use of Monte Carlo methods in the calculation of an overall probability distribution and presents a logical and mathematical foundation for use of the scenario concept in such calculations. The report also draws preliminary conclusions regarding the shape of the probability distribution for the WIPP system; preliminary conclusions are based on the possible occurrence of three events and the presence of one feature: namely, the events ''attempted boreholes over rooms and drifts,'' ''mining alters ground-water regime,'' ''water-withdrawal wells provide alternate pathways,'' and the feature ''brine pocket below room or drift.'' Calculation of the WIPP systems's overall probability distributions for only five of sixteen possible scenario classes that can be obtained by combining the four postulated events or features

Preliminary Modeling Of Radiation Levels At The Fermilab PIP-II Linac arXiv

CERN Document Server

Lari, L.; Esposito, L.S.; Baffes, C.; Dixon, S.J.; Mokhov, N.V.; Rakhno, I.; Tropin, I.S.

PIP-II is the Fermilab's flagship project for providing powerful, high-intensity proton beams to the laboratory's experiments. The heart of PIP-II is an 800-MeV superconducting linac accelerator. It will be located in a new tunnel with new service buildings and connected to the present Booster through a new transfer line. To support the design of civil engineering and mechanical integration, this paper provides preliminary estimation of radiation level in the gallery at an operational beam loss limit of 0.1 W/m, by means of Monte Carlo calculations with FLUKA and MARS15 codes.

Preliminary Computational Fluid Dynamics (CFD) Simulation of EIIB Push Barge in Shallow Water

Science.gov (United States)

Beneš, Petr; Kollárik, Róbert

2011-12-01

This study presents preliminary CFD simulation of EIIb push barge in inland conditions using CFD software Ansys Fluent. The RANSE (Reynolds Averaged Navier-Stokes Equation) methods are used for the viscosity solution of turbulent flow around the ship hull. Different RANSE methods are used for the comparison of their results in ship resistance calculations, for selecting the appropriate and removing inappropriate methods. This study further familiarizes on the creation of geometrical model which considers exact water depth to vessel draft ratio in shallow water conditions, grid generation, setting mathematical model in Fluent and evaluation of the simulations results.

Preliminary estimates of cost savings for defense high level waste vitrification options

International Nuclear Information System (INIS)

Merrill, R.A.; Chapman, C.C.

1993-09-01

The potential for realizing cost savings in the disposal of defense high-level waste through process and design modificatins has been considered. Proposed modifications range from simple changes in the canister design to development of an advanced melter capable of processing glass with a higher waste loading. Preliminary calculations estimate the total disposal cost (not including capital or operating costs) for defense high-level waste to be about $7.9 billion dollars for the reference conditions described in this paper, while projected savings resulting from the proposed process and design changes could reduce the disposal cost of defense high-level waste by up to $5.2 billion

Results of questionnaire for the needs of measured data for the steady-state calculations

International Nuclear Information System (INIS)

Yrjoelae, V.

1995-01-01

In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given

Results of questionnaire for the needs of measured data for the steady-state calculations

Energy Technology Data Exchange (ETDEWEB)

Yrjoelae, V. [VTT Energy, Espoo (Finland)

1995-12-31

In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given.

Results of questionnaire for the needs of measured data for the steady-state calculations

Energy Technology Data Exchange (ETDEWEB)

Yrjoelae, V [VTT Energy, Espoo (Finland)

1996-12-31

In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given.

Preliminary analysis for model development of groundwater evolution in Horonobe area

International Nuclear Information System (INIS)

Yoshida, Yasushi; Yui, Mikazu

2003-03-01

The preliminary analysis for model development of groundwater evolution in Horonobe area was performed with data at D-1, HDB-1 and HDB-2 bore hole where hydrogen / oxygen isotope concentration, mineral property in sedimentary rock and physico-chemical parameters (pH, Eh and ionic concentrations) were measured. As a result of analysis for hydrogen and oxygen isotope concentration, saline water in marine sediment was diluted by the mixing with shallow groundwater and diffusion. And as a result of analysis for a concentration of bicarbonate from the difference of pH values measured between in-situ and under air, the estimated in-situ concentration of bicarbonate differs from that measured under air. And minerals which were assumed to be equilibrium with groundwater were selected by thermodynamic calculation. This report presents the results of preliminary analysis for groundwater evolution by using data derived from D-1, HDB-1 and HDB-2 boring research. In order to establish the model which interprets the groundwater evolution as a next step, data which satisfy the representative in-situ property of groundwater chemistry in Horonobe area are needed. Reliable measurements for physico-chemical parameter and property of minerals in sedimentary rock in dominant layer and at the variety of depth are also needed. (author)

Exploratory shaft facility preliminary designs - Gulf Interior Region salt domes

International Nuclear Information System (INIS)

1983-09-01

The purpose of the Preliminary Design Report, Gulf Interior Region, is to provide a description of the preliminary design for an Exploratory Shaft Facility on the Richton Dome, Mississippi. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers are included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description and Construction Cost Estimate

A preliminary conceptual design study for Korean fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

2013-10-15

Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

Purification, crystallization and preliminary X-ray study of the fungal laccase from Cerrena maxima

International Nuclear Information System (INIS)

Lyashenko, Andrey V.; Zhukhlistova, Nadegda E.; Gabdoulkhakov, Azat G.; Zhukova, Yuliya N.; Voelter, Wolfang; Zaitsev, Viatcheslav N.; Bento, Isabel; Stepanova, Elena V.; Kachalova, Galina S.; Koroleva, Ol’ga V.; Cherkashyn, Evgeniy A.; Tishkov, Vladimir I.; Lamzin, Victor S.; Schirwitz, Katja; Morgunova, Ekaterina Yu.; Betzel, Christian; Lindley, Peter F.; Mikhailov, Al’bert M.

2006-01-01

The crystallization and preliminary X-ray structure at 1.9 Å resolution of the fungal laccase from C. maxima are presented. Laccases are members of the blue multi-copper oxidase family that oxidize substrate molecules by accepting electrons at a mononuclear copper centre and transferring them to a trinuclear centre. Dioxygen binds to the trinuclear centre and, following the transfer of four electrons, is reduced to two molecules of water. Crystals of the laccase from Cerrena maxima have been obtained and X-ray data were collected to 1.9 Å resolution using synchrotron radiation. A preliminary analysis shows that the enzyme has the typical laccase structure and several carbohydrate sites have been identified. The carbohydrate chains appear to be involved in stabilization of the intermolecular contacts in the crystal structure, thus promoting the formation of well ordered crystals of the enzyme. Here, the results of an X-ray crystallographic study on the laccase from the fungus Cerrena maxima are reported. Crystals that diffract well to a resolution of at least 1.9 Å (R factor = 18.953%; R free = 23.835; r.m.s.d. bond lengths, 0.06 Å; r.m.s.d. bond angles, 1.07°) have been obtained despite the presence of glycan moieties. The overall spatial organization of C. maxima laccase and the structure of its copper-containing active centre have been determined by the molecular-replacement method using the laccase from Trametes versicolor (Piontek et al., 2002 ▶) as a structural template. In addition, four glycan-binding sites were identified and the 1.9 Å X-ray data were used to determine the previously unknown primary structure of this protein. The identity (calculated from sequence alignment) between the C. maxima laccase and the T. versicolor laccase is about 87%. Tyr196 and Tyr372 show significant extra density at the ortho positions and this has been interpreted in terms of NO 2 substituents

(11)C-labeling and preliminary evaluation of vortioxetine as a PET radioligand

DEFF Research Database (Denmark)

Andersen, Valdemar L; Hansen, Hanne D; Herth, Matthias M

2014-01-01

. Preliminary evaluation of [(11)C]vortioxetine in a Danish Landrace pig showed rapid brain uptake and brain distribution in accordance with the pharmacological profile, all though an unexpected high binding in cerebellum was also observed. [(11)C]vortioxetine displayed slow tracer kinetics with peak uptake...

Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

2003-04-01

In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed.

Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

2003-04-01

In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed

GTHTR 300 economic calculation with Mini G4ECONS as a basis for generation cost of GTHTR 10 MWe calculation

International Nuclear Information System (INIS)

Mochamad Nasrullah; Nurlaila

2014-01-01

The government plan to build Experimental Power Reactor (EPR) requires measurable economic assessment. The purpose of the study was to recalculate Gas Turbine High Temperature Reactor of 300 MWe (GTHTR 300) and compare the results with reference data. Then calculate generation cost of GTHTR 3, 5 and 10 MWe using the scale factor calculation. The methodology used is covered the generation cost calculation using the Mini G4Econs spread sheet models published by IAEA. Result of the verification calculation showed that a relatively similar, which means that the calculation model could be used to calculate for same other cases. Afterward, using scale factor, smaller scale reactor could be calculated. The calculation result show that electricity generation cost of SMR-HTR type with load factor 90% and discount rate 10% for power capacity 3, 5 and 10 MWe are 29.5, 22.68 and 16.17 cents$/kWh respectively. However, because the EPR is planning to be built as a non-commercial power reactors, then 5 % and 3 % of discount rate could be chosen, each of those discount rate will result electricity generation cost of 10.37 cents$/kWh and 8.56 cents$/kWh respectively. These result could be considered by the government for developing SMR type of HTR. (author)

75 FR 984 - Draft Recommended Interim Preliminary Remediation Goals for Dioxin in Soil at CERCLA and RCRA Sites

Science.gov (United States)

2010-01-07

...The Environmental Protection Agency (EPA or the Agency) is announcing a 50-day public comment period for draft recommended interim preliminary remediation goals (PRGs) developed in the Draft Recommended Interim Preliminary Remediation Goals for Dioxin in Soil at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and Resource Conservation and Recovery Act (RCRA) Sites. EPA's Office of Solid Waste and Emergency and Emergency Response (OSWER) has developed the draft recommended interim PRGs for dioxin in soil. These draft recommended interim PRGs were calculated using existing, peer- reviewed toxicity values and current EPA equations and default exposure assumptions. This Federal Register notice is intended to provide an opportunity for public comment on the draft recommended interim PRGs. EPA will consider any public comments submitted in accordance with this notice and may revise the draft recommended interim PRGs thereafter.

The MiAge Calculator: a DNA methylation-based mitotic age calculator of human tissue types.

Science.gov (United States)

Youn, Ahrim; Wang, Shuang

2018-01-01

Cell division is important in human aging and cancer. The estimation of the number of cell divisions (mitotic age) of a given tissue type in individuals is of great interest as it allows not only the study of biological aging (using a new molecular aging target) but also the stratification of prospective cancer risk. Here, we introduce the MiAge Calculator, a mitotic age calculator based on a novel statistical framework, the MiAge model. MiAge is designed to quantitatively estimate mitotic age (total number of lifetime cell divisions) of a tissue using the stochastic replication errors accumulated in the epigenetic inheritance process during cell divisions. With the MiAge model, the MiAge Calculator was built using the training data of DNA methylation measures of 4,020 tumor and adjacent normal tissue samples from eight TCGA cancer types and was tested using the testing data of DNA methylation measures of 2,221 tumor and adjacent normal tissue samples of five other TCGA cancer types. We showed that within each of the thirteen cancer types studied, the estimated mitotic age is universally accelerated in tumor tissues compared to adjacent normal tissues. Across the thirteen cancer types, we showed that worse cancer survivals are associated with more accelerated mitotic age in tumor tissues. Importantly, we demonstrated the utility of mitotic age by showing that the integration of mitotic age and clinical information leads to improved survival prediction in six out of the thirteen cancer types studied. The MiAge Calculator is available at http://www.columbia.edu/∼sw2206/softwares.htm .

Two-dimensional computer simulation of hypervelocity impact cratering: some preliminary results for Meteor Crater, Arizona

International Nuclear Information System (INIS)

Bryan, J.B.; Burton, D.E.; Cunningham, M.E.; Lettis, L.A. Jr.

1978-04-01

A computational approach used for subsurface explosion cratering has been extended to hypervelocity impact cratering. Meteor (Barringer) Crater, Arizona, was selected for our first computer simulation because it was the most thoroughly studied. It is also an excellent example of a simple, bowl-shaped crater and is one of the youngest terrestrial impact craters. Shoemaker estimates that the impact occurred about 20,000 to 30,000 years ago [Roddy (1977)]. Initial conditions for this calculation included a meteorite impact velocity of 15 km/s. meteorite mass of 1.57E + 08 kg, with a corresponding kinetic energy of 1.88E + 16 J (4.5 megatons). A two-dimensional Eulerian finite difference code called SOIL was used for this simulation of a cylindrical iron projectile impacting at normal incidence into a limestone target. For this initial calculation a Tillotson equation-of-state description for iron and limestone was used with no shear strength. A color movie based on this calculation was produced using computer-generated graphics. Results obtained for this preliminary calculation of the formation of Meteor Crater, Arizona, are in good agreement with Meteor Crater Measurements

Benchmark calculations with simple phantom for neutron dosimetry (2)

International Nuclear Information System (INIS)

Yukio, Sakamoto; Shuichi, Tsuda; Tatsuhiko, Sato; Nobuaki, Yoshizawa; Hideo, Hirayama

2004-01-01

Benchmark calculations for high-energy neutron dosimetry were undertaken after SATIF-5. Energy deposition in a cylindrical phantom with 100 cm radius and 30 cm depth was calculated for the irradiation of neutrons from 100 MeV to 10 GeV. Using the ICRU four-element loft tissue phantom and four single-element (hydrogen, carbon, nitrogen and oxygen) phantoms, the depth distributions of deposition energy and those total at the central region of phantoms within l cm radius and at the whole region of phantoms within 100 cm radius were calculated. The calculated results of FLUKA, MCNPX, MARS, HETC-3STEP and NMTC/JAM codes were compared. It was found that FLUKA, MARS and NMTC/JAM showed almost the same results. For the high-energy neutron incident, the MCNP-X results showed the largest ones in the total deposition energy and the HETC-3STEP results show'ed smallest ones. (author)

Critical and subcritical mass calculations of fissionable nuclides based on JENDL-3.2+

International Nuclear Information System (INIS)

Okuno, H.

2002-01-01

We calculated critical and subcritical masses of 10 fissionable actinides ( 233 U, 235 U, 238 Pu, 239 Pu, 241 Pu, 242m Am, 243 Cm, 244 Cm, 249 Cf and 251 Cf) in metal and in metal-water mixtures (except 238 Pu and 244 Cm). The calculation was made with a combination of a continuous energy Monte Carlo neutron transport code, MCNP-4B2, and the latest released version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI, JEF-2.2, and JENDL-3.3 in its preliminary version were also applied to find differences in results originated from different nuclear data files. For the so-called big three fissiles ( 233 U, 235 U and 239 Pu), analyzing the criticality experiments cited in ICSBEP Handbook validated the code-library combination, and calculation errors were consequently evaluated. Estimated critical and lower limit critical masses of the big three in a sphere with/without a water or SS-304 reflector were supplied, and they were compared with the subcritical mass limits of ANS-8.1. (author)

Preliminary results of the seventh three-dimensional AER dynamic benchmark problem calculation. Solution with DYN3D and RELAP5-3D codes

International Nuclear Information System (INIS)

Bencik, M.; Hadek, J.

2011-01-01

The paper gives a brief survey of the seventh three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at Nuclear Research Institute Rez. This benchmark was defined at the twentieth AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a WWER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The WWER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D. The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the seventh AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. (Authors)

Comments on SKB's SFL 3-5 preliminary performance assessment

International Nuclear Information System (INIS)

Wilmot, R.D.; Crawford, M.B.

2000-01-01

Recently introduced regulations in Sweden have established an individual risk criterion ( -6 per year) for the long-term performance of repositories for the disposal of radioactive wastes. SKB has not focused its assessment of SFL 3-5 on demonstrating compliance with this regulation. Instead, SKB has calculated individual dose and provided a comparison with an annual individual dose of 14 iSv (derived from the risk criteria using the ICRP's dose-risk conversion factor of 0.073 per Sv). The justification of this approach is that probabilities do not need to be determined if doses are less than the dose equivalent to the risk criterion. However, there is insufficient information regarding uncertainty provided in the documentation of the SFL 3-5 assessment to determine whether this approach is reasonable. SKB's parallel assessment of a repository for spent fuel using the KBS-3 concept (SR 97) accounts for uncertainty by specifying a 'reasonable' and a 'pessimistic' value for uncertain parameters in the assessment calculations. Although there are problems with the way probabilities have been assigned to these values, this approach does indicate where there are significant uncertainties. The SFL 3-5 PA does not include a structured approach to defining uncertainty, although a number of assumptions and parameter values are stated to be conservative. As a preliminary assessment, there is insufficient information to identify key uncertainties or sensitivities, or to determine where further work should be focused. Any assessment requires the use of expert judgement to determine how the assessment is conducted, what modelling approach to use, what features, events and processes (FEPs) could potentially affect the disposal system, which FEPs should be included in the conceptual models, and which scenarios should be assessed. Judgements are also required in determining how to parameterize the models, and this may extend to formal expert elicitation for particular parameter

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

Directory of Open Access Journals (Sweden)

Porta A.

2016-01-01

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

Translocation of 11C from leaves of Helianthus: preliminary results

International Nuclear Information System (INIS)

Fensom, D.S.; Aikman, D.; Scobie, J.; Drinkwater, A.; Ledingham, K.W.O.

1977-01-01

11 C fed to leaves as 11 CO 2 was used to study the dynamics of short-term translocation of photosynthate in Helianthus. As in 14 C studies small amounts of tracer were often detected in the stem close to the fed leaf in th first 5 min, followed by a larger mass flow after 15 min. The speed of mass flow of tracer movement was calculated to be 60 to 400 cm.h -1 depending on the method of calculation. There was no evidence in the premass flow for discrete spots along the stem or petiole where tracer accumulated. Neither was there firm evidence for pulses of tracer moving steadily forward, but there were point fluctuations of greater variability than would be expected by chance alone, which suggest the possibility of aberrations of movement superimposed on the mass flow. Details of these aberrations could not be assessed with certainty from these preliminary experiments owing to the rather low tracer activity. The translocation profiles were sensitive to the prior light conditioning of the plant and above all to chilling. In Helianthus the latter produced temporary restrictions in translocation which lasted for some 10-12 min. (author)

46 CFR 176.635 - Preliminary examination requirements.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Preliminary examination requirements. 176.635 Section... (UNDER 100 GROSS TONS) INSPECTION AND CERTIFICATION Hull and Tailshaft Examinations § 176.635 Preliminary examination requirements. (a) If you exclusively use divers to examine the underwater hull plating, you must...

Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

International Nuclear Information System (INIS)

Mozzillo, Rocco; Tarallo, Andrea; Marzullo, Domenico; Bachmann, Christian; Di Gironimo, Giuseppe; Mazzone, Giuseppe

2016-01-01

Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

Energy Technology Data Exchange (ETDEWEB)

Mozzillo, Rocco, E-mail: rocco.mozzillo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Tarallo, Andrea; Marzullo, Domenico [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione - ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

2016-11-15

Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

Scope and purpose of the preliminary planning work

International Nuclear Information System (INIS)

Kalas, P.

1976-01-01

The results of preliminary planning work are usually expressed in a number of recommendations covering mainly: long-term national policy in the field of energy resources and selection of projects to be further studied at the feasibility level. Moreover, recommendations on further actions are made including: inventory of generation and transmission facilities recommended for the implementation in order to meet the load forecasted for medium-term period, preparation of a preliminary calender of decisions to be taken for the implementation of the projects recommended, preparation of a preliminary construction schedule, preparation of a preliminary investment program, preparation of a program of necessary engineering works, and performance of study on electricity rates which would adjust existing tariffs to proposed development program of the utility. (HP) [de

Helical tomotherapy shielding calculation for an existing LINAC treatment room: sample calculation and cautions

International Nuclear Information System (INIS)

Wu Chuan; Guo Fanqing; Purdy, James A

2006-01-01

This paper reports a step-by-step shielding calculation recipe for a helical tomotherapy unit (TomoTherapy Inc., Madison, WI, USA), recently installed in an existing Varian 600C treatment room. Both primary and secondary radiations (leakage and scatter) are explicitly considered. A typical patient load is assumed. Use factor is calculated based on an analytical formula derived from the tomotherapy rotational beam delivery geometry. Leakage and scatter are included in the calculation based on corresponding measurement data as documented by TomoTherapy Inc. Our calculation result shows that, except for a small area by the therapists' console, most of the existing Varian 600C shielding is sufficient for the new tomotherapy unit. This work cautions other institutions facing the similar situation, where an HT unit is considered for an existing LINAC treatment room, more secondary shielding might be considered at some locations, due to the significantly increased secondary shielding requirement by HT. (note)

Preliminary results of steel containment vessel model test

International Nuclear Information System (INIS)

Matsumoto, T.; Komine, K.; Arai, S.

1997-01-01

A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11-12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented

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)

Calculation for simulation of archery goal value using a web camera and ultrasonic sensor

Science.gov (United States)

Rusjdi, Darma; Abdurrasyid, Wulandari, Dewi Arianti

2017-08-01

Development of the device simulator digital indoor archery-based embedded systems as a solution to the limitations of the field or open space is adequate, especially in big cities. Development of the device requires simulations to calculate the value of achieving the target based on the approach defined by the parabolic motion variable initial velocity and direction of motion of the arrow reaches the target. The simulator device should be complemented with an initial velocity measuring device using ultrasonic sensors and measuring direction of the target using a digital camera. The methodology uses research and development of application software from modeling and simulation approach. The research objective to create simulation applications calculating the value of the achievement of the target arrows. Benefits as a preliminary stage for the development of the simulator device of archery. Implementation of calculating the value of the target arrows into the application program generates a simulation game of archery that can be used as a reference development of the digital archery simulator in a room with embedded systems using ultrasonic sensors and web cameras. Applications developed with the simulation calculation comparing the outer radius of the circle produced a camera from a distance of three meters.

Teletandem between French and Brazilian students: Some preliminary remarks

Directory of Open Access Journals (Sweden)

Liliane SANTOS

2015-12-01

Full Text Available ABSTRACT In its French-Brazilian version, the Teletandem Brazil project enables students from the University of Lille 3 (France and from the State University of São Paulo (Unesp, Brazil, to take part in online exchanges, based on the principles of autonomy and reciprocity. In this work, we will present some preliminary remarks on the construction of cultural identity representations by the students who took part in the project, from 2006 to 2012, the specificity of the exchanges we analyze being that most of the French students involved in them are third generation Portuguese. We will examine the consequences of the introduction of a third culture within exchanges which, linguistically speaking, are bilateral. The French students are often experiencing conflicting feelings toward Brazil and, similarly, the Brazilian students may have conflicting feelings towards Portugal and France. Our preliminary results show that the most successful linguistic exchanges occur when students face their own cultural identity with no feeling of superiority or inferiority.

Preliminary geologic map of the Lathrop Wells volcanic center

International Nuclear Information System (INIS)

Crowe, B.; Harrington, C.; McFadden, L.; Perry, F.; Wells, S.; Turrin, B.; Champion, D.

1988-12-01

A preliminary geologic map has been compiled for the bedrock geology of the Lathrop Wells volcanic center. The map was completed through use of a combination of stereo photographic interpretation and field mapping on color aerial photographs. These photographs (scale 1:4000) were obtained from American Aerial Surveys, Inc. They were flown on August 18, 1987, at the request of the Yucca Mountain Project (then Nevada Nuclear Waste Storage Investigations). The photographs are the Lathrop Wells VC-Area 25 series, numbers 1--32. The original negatives for these photographs are on file with American Aerial Surveys, Inc. Copies of the negatives have been archived at the Los Alamos National Laboratory, Group N-5. The preliminary geologic map is a bedrock geologic map. It does not show alluvial deposits, eolian sands, or scoria fall deposits from the youngest eruptive events. The units will be compiled on separate maps when the geomorphic and soils studies are more advanced

External radiotherapy in macular degeneration: Our technique, dosimetric calculation, and preliminary results

International Nuclear Information System (INIS)

Akmansu, M.; Dirican, Bahar; Oeztuerk, Berrin; Egehan, Ibrahim; Subasi, Mahmut; Or, Meral

1998-01-01

Purpose: This study was performed to determine the toxicity and efficacy of external-beam radiotherapy in patients with age-related subfoveal neovascularization. Methods and Materials: Between January 1996 and September 1996, 25 patients with a mean age of 70.5 (60-84) years were enrolled. All patients underwent fluorescein angiographic evaluation and documentation of their neovascular disease prior to irradiation. A total of 25 patients were treated with a total dose of 12 Gy in 6 fractions over 8 days. We used a lens-sparing technique and patients were treated with a single lateral 6-MV photon beam. To assess the risk of radiation carcinogenesis after treatment of age-related subfoveal neovascularization, we estimated the effective dose for a standard patient on the basis of tissue-weighting factors as defined by the International Commission on Radiological Protection (ICRP). The calculations were made with TLD on a male randophantom. The lens dose was found to be 0.217 Gy per fraction. Results: No significant acute morbidity was noted. Visual acuity was maintained or improved in 76% and 80% of treated patients at their 1- and 3-month follow-up examinations, respectively. On angiographic imaging, there was stabilization of subfoveal neovascular membranes in 23 patients (92%) at 3 months after irradiation. Conclusion: Our observations on these 25 patients in this study indicate that many patients will have improved or stable vision after radiotherapy treatment with low-dose irradiation

Preliminary aseismic analysis on bolts of driving mechanism in absorption sphere shutdown system

International Nuclear Information System (INIS)

Chen Feng; Li Tianjin; Zhang Zhengming; Huang Zhiyong; Bo Hanliang

2012-01-01

The absorption sphere shutdown system performs an important role in reactivity regulating and control. Driving mechanism is a set of key mechanical moving parts which is used to control falling of absorption spheres in absorption sphere shutdown system. It is about 5 m for driving mechanism with the slim structure, which is connected with the upper supported plate of metal reactor internals through storage vessel with bolts. Both the storage vessel and driving mechanism are equipment of seismic classification I. It is significant to calculate and check the bolts strength of driving mechanism. In this paper, complicate structure of driving mechanism was simplified to three variable cross sections and statically indeterminate problem was solved. The bolts at the bottom and on the top of the storage vessel were calculated and checked. The preliminary results indicate that the bolts strength is reliable and safe, and the supporting force at the most weak point of driving mechanism is as well obtained. (authors)

Preliminary calculation with code CONTEMPT-LT for spray cooling tests with JAERI model containment vessel

International Nuclear Information System (INIS)

Tanaka, Mitsugu

1978-01-01

LWR plants have a containment spray system to reduce the escape of radioactive material to the environment in a loss-of-coolant accident (LOCA) by washing out fission products, especially radioiodine, and condensing the steam to lower the pressure. For carrying out the containment spray tests, pressure and temperature behaviour of the JAERI Model Containment Vessel in spray cooling has been calculated with computer program CONTEMPT-LT. The following could be studied quantitatively: (1) pressure and temperature raise rates for steam addition rate and (2) pressure fall rate for spray flow rate and spray heat transfer efficiency. (auth.)

Mordred: a molecular descriptor calculator.

Science.gov (United States)

Moriwaki, Hirotomo; Tian, Yu-Shi; Kawashita, Norihito; Takagi, Tatsuya

2018-02-06

Molecular descriptors are widely employed to present molecular characteristics in cheminformatics. Various molecular-descriptor-calculation software programs have been developed. However, users of those programs must contend with several issues, including software bugs, insufficient update frequencies, and software licensing constraints. To address these issues, we propose Mordred, a developed descriptor-calculation software application that can calculate more than 1800 two- and three-dimensional descriptors. It is freely available via GitHub. Mordred can be easily installed and used in the command line interface, as a web application, or as a high-flexibility Python package on all major platforms (Windows, Linux, and macOS). Performance benchmark results show that Mordred is at least twice as fast as the well-known PaDEL-Descriptor and it can calculate descriptors for large molecules, which cannot be accomplished by other software. Owing to its good performance, convenience, number of descriptors, and a lax licensing constraint, Mordred is a promising choice of molecular descriptor calculation software that can be utilized for cheminformatics studies, such as those on quantitative structure-property relationships.

Preliminary dose assessment of the Chernobyl accident

International Nuclear Information System (INIS)

Hull, A.P.

1987-01-01

From the major accident at Unit 4 of the Chernobyl nuclear power station, a plume of airborne radioactive fission products was initially carried northwesterly toward Poland, thence toward Scandinavia and into Central Europe. Reports of the levels of radioactivity in a variety of media and of external radiation levels were collected in the Department of Energy's Emergency Operations Center and compiled into a data bank. Portions of these and other data which were obtained directly from published and official reports were utilized to make a preliminary assessment of the extent and magnitude of the external dose to individuals downwind from Chernobyl. Radioactive 131 I was the predominant fission product. The time of arrival of the plume and the maximum concentrations of 131 I in air, vegetation and milk and the maximum reported depositions and external radiation levels have been tabulated country by country. A large amount of the total activity in the release was apparently carried to a significant elevation. The data suggest that in areas where rainfall occurred, deposition levels were from ten to one-hundred times those observed in nearby ''dry'' locations. Sufficient spectral data were obtained to establish average release fractions and to establish a reference spectra of the other nuclides in the release. Preliminary calculations indicated that the collective dose equivalent to the population in Scandinavia and Central Europe during the first year after the Chernobyl accident would be about 8 x 10 6 person-rem. From the Soviet report, it appears that a first year population dose of about 2 x 10 7 person-rem (2 x 10 5 Sv) will be received by the population who were downwind of Chernobyl within the U.S.S.R. during the accident and its subsequent releases over the following week. 32 refs., 14 figs., 20 tabs

7 CFR 1780.55 - Preliminary engineering reports and Environmental Reports.

Science.gov (United States)

2010-01-01

... 7 Agriculture 12 2010-01-01 2010-01-01 false Preliminary engineering reports and Environmental..., Designing, Bidding, Contracting, Constructing and Inspections § 1780.55 Preliminary engineering reports and Environmental Reports. Preliminary engineering reports (PERs) must conform to customary professional standards...

Preliminary Context Analysis of Community Informatics Social ...

African Journals Online (AJOL)

Preliminary context analysis is always part of the feasibility study phase in the development of information system for Community Development (CD) purposes. In this paper, a context model and a preliminary context analysis are presented for Social Network Web Application (SNWA) for CD in the Niger Delta region of ...

Self-field calculation of CICC with fast direct Biot–Savart integration

Energy Technology Data Exchange (ETDEWEB)

Wang, Xu; Li, Yingxu [Key Laboratory of Mechanics on Environment and Disaster in Western China, The Ministry of Education of China, Lanzhou, Gansu 730000 (China); Department of Mechanics and Engineering Science, College of Civil Engineering and Mechanics, Lanzhou University, Lanzhou, Gansu 730000 (China); Gao, Yuanwen, E-mail: ywgao@lzu.edu.cn [Key Laboratory of Mechanics on Environment and Disaster in Western China, The Ministry of Education of China, Lanzhou, Gansu 730000 (China); Department of Mechanics and Engineering Science, College of Civil Engineering and Mechanics, Lanzhou University, Lanzhou, Gansu 730000 (China); Zhou, Youhe [Key Laboratory of Mechanics on Environment and Disaster in Western China, The Ministry of Education of China, Lanzhou, Gansu 730000 (China); Department of Mechanics and Engineering Science, College of Civil Engineering and Mechanics, Lanzhou University, Lanzhou, Gansu 730000 (China)

2014-04-15

Highlights: • An algorithm of fast direct Biot–Savart integration (FDBS) is proposed. • FDBS calculates the self-field of ITER cable-in-conduit conductor (CICC). • FDBS is more effective and easier to implement. • This new method will benefit future magnet design. - Abstract: ITER magnetic device (Tokamak) requires a strong magnetic field produced by charged cable conductors and external sources to arrive at stable and reliable magnetic confinement performance. Before manufacturing and assembling conductors, preliminary analysis of self-field induction is helpful for reducing the cost of varying-parameter experiments. Spatial helix shape of numerous strand elements and multi-level twist of the finalized cable, known as CICC type, make it unpractical to direct use finite-element methods and other numerical procedures for self-field calculation. An algorithm FDBS (fast direct Biot–Savart integration) is proposed to surmount this difficulty, which improves the traditional method (DBS, direct implementing Biot–Savart law for all strand sources) in terms of computational effort. As such the complexity reduces to O(N) from the original O(N{sup 2}) and speed enhancement is achieved in the parallel computation environment. FDBS calculates out a detailed self-field profile for the uncompressed ITER TF conductors carrying uniform current at each cabling level; the layered self-field distribution becomes more indistinct for higher level subcable.

Self-field calculation of CICC with fast direct Biot–Savart integration

International Nuclear Information System (INIS)

Wang, Xu; Li, Yingxu; Gao, Yuanwen; Zhou, Youhe

2014-01-01

Highlights: • An algorithm of fast direct Biot–Savart integration (FDBS) is proposed. • FDBS calculates the self-field of ITER cable-in-conduit conductor (CICC). • FDBS is more effective and easier to implement. • This new method will benefit future magnet design. - Abstract: ITER magnetic device (Tokamak) requires a strong magnetic field produced by charged cable conductors and external sources to arrive at stable and reliable magnetic confinement performance. Before manufacturing and assembling conductors, preliminary analysis of self-field induction is helpful for reducing the cost of varying-parameter experiments. Spatial helix shape of numerous strand elements and multi-level twist of the finalized cable, known as CICC type, make it unpractical to direct use finite-element methods and other numerical procedures for self-field calculation. An algorithm FDBS (fast direct Biot–Savart integration) is proposed to surmount this difficulty, which improves the traditional method (DBS, direct implementing Biot–Savart law for all strand sources) in terms of computational effort. As such the complexity reduces to O(N) from the original O(N 2 ) and speed enhancement is achieved in the parallel computation environment. FDBS calculates out a detailed self-field profile for the uncompressed ITER TF conductors carrying uniform current at each cabling level; the layered self-field distribution becomes more indistinct for higher level subcable

Some calculator programs for particle physics

International Nuclear Information System (INIS)

Wohl, C.G.

1982-01-01

Seven calculator programs that do simple chores that arise in elementary particle physics are given. LEGENDRE evaluates the Legendre polynomial series Σa/sub n/P/sub n/(x) at a series of values of x. ASSOCIATED LEGENDRE evaluates the first-associated Legendre polynomial series Σb/sub n/P/sub n/ 1 (x) at a series of values of x. CONFIDENCE calculates confidence levels for chi 2 , Gaussian, or Poisson probability distributions. TWO BODY calculates the c.m. energy, the initial- and final-state c.m. momenta, and the extreme values of t and u for a 2-body reaction. ELLIPSE calculates coordinates of points for drawing an ellipse plot showing the kinematics of a 2-body reaction or decay. DALITZ RECTANGULAR calculates coordinates of points on the boundary of a rectangular Dalitz plot. DALITZ TRIANGULAR calculates coordinates of points on the boundary of a triangular Dalitz plot. There are short versions of CONFIDENCE (EVEN N and POISSON) that calculate confidence levels for the even-degree-of-freedom-chi 2 and the Poisson cases, and there is a short version of TWO BODY (CM) that calculates just the c.m. energy and initial-state momentum. The programs are written for the HP-97 calculator

Purification, crystallization and preliminary X-ray diffraction studies of parakeet (Psittacula krameri) haemoglobin.

Science.gov (United States)

Jaimohan, S M; Naresh, M D; Arumugam, V; Mandal, A B

2009-10-01

Birds often show efficient oxygen management in order to meet the special demands of their metabolism. However, the structural studies of avian haemoglobins (Hbs) are inadequate for complete understanding of the mechanism involved. Towards this end, purification, crystallization and preliminary X-ray diffraction studies have been carried out for parakeet Hb. Parakeet Hb was crystallized as the met form in low-salt buffered conditions after extracting haemoglobin from crude blood by microcentrifugation and purifying the sample by column chromatography. Good-quality crystals were grown from 10% PEG 3350 and a crystal diffracted to about 2.8 A resolution. Preliminary diffraction data showed that the Hb crystal belonged to the monoclinic system (space group C2), with unit-cell parameters a = 110.68, b = 64.27, c = 56.40 A, beta = 109.35 degrees . Matthews volume analysis indicated that the crystals contained a half-tetramer in the asymmetric unit.

An Assessment of SKB's Performance Assessment Calculations in the Interim Main Report for the Safety Assessment SR-Can

International Nuclear Information System (INIS)

Maul, Philip; Robinson, Peter

2005-03-01

presented. It is suggested that in the SR-Can safety assessment sufficient information should be provided to enable PA calculations to be fully reproduced. 9. SKB's preliminary calculations indicate that safety criteria are likely to be met by a comfortable margin even if a large fraction of the canisters fail. The geosphere plays only a minor role in the retardation function. This assertion depends on a number of assumptions, and the independent AMBER calculations show that different assumptions could mean that the safety criteria would be exceeded in this extreme case

Bridging quantum chemistry and nuclear structure theory: Coupled-cluster calculations for closed- and open-shell nuclei

International Nuclear Information System (INIS)

Piecuch, Piotr; Wloch, Marta; Gour, Jeffrey R.; Dean, David J.; Papenbrock, Thomas; Hjorth-Jensen, Morten

2005-01-01

We review basic elements of the single-reference coupled-cluster theory and discuss large scale ab initio calculations of ground and excited states of 15O, 16O, and 17O using coupled-cluster methods and algorithms developed in quantum chemistry. By using realistic two-body interactions and the renormalized form of the Hamiltonian obtained with a no-core G-matrix approach, we obtain the converged results for 16O and promising preliminary results for 15O and 17O at the level of two-body interactions. The calculated properties other than energies include matter density, charge radius, and charge form factor. The relatively low costs of coupled-cluster calculations, which are characterized by the low-order polynomial scaling with the system size, enable us to probe large model spaces with up to 7 or 8 major oscillator shells, for which non-truncated shell-model calculations for nuclei with A = 15 17 active particles are presently not possible. We argue that the use of coupled-cluster methods and computer algorithms developed by quantum chemists to calculate properties of nuclei is an important step toward the development of accurate and affordable many-body theories that cross the boundaries of various physical sciences

Preliminary comparative estimate of the environmental externalities of the electrical generation in Cuba

International Nuclear Information System (INIS)

Turtos Carbonell, L.

1998-01-01

Determination of the externalises associated with the electrical generation and fundamentally its atmospherically environmental impact, win greater importance nowadays, with the objective that to medium term these could be incorporated into the economy of electricity production as the surest way to reduce this impact. In the work is accomplished a comparative preliminary estimate of the externalises of the electrical generation in Cuba based in the results obtained in the External Project (Externalises of Energy) and the emissions of the domestic Power Plant. Different processes to reduce these emissions are proposed. The economic feasibility of installing Abatement Emissions Technologies based on the calculated externalises is analyzed

National Courts of Last Instance Failing to Make a Preliminary Reference

DEFF Research Database (Denmark)

Broberg, Morten

2016-01-01

are the consequences if a Member State court fails to make a preliminary reference in a situation where it was legally obliged to do so? The article shows that such failure may constitute an infringement of the right to a fair trial as laid down in Article 6(1) of the European Convention of Human Rights. It may also...

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.

Spreadsheet eases heat balance, payback calculations

International Nuclear Information System (INIS)

Conner, K.P.

1992-01-01

This paper reports that a generalized Lotus type spreadsheet program has been developed to perform the heat balance and simple payback calculations for various turbine-generator (TG) inlet steam pressures. It can be used for potential plant expansions or new cogeneration installations. The program performs the basic heat balance calculations that are associated with turbine-generator, feedwater heating process steam requirements and desuperheating. The printout, shows the basic data and formulation used in the calculations. The turbine efficiency data used are applicable for automatic extraction turbine-generators in the 30-80 MW range. Simple payback calculations are for chemical recovery boilers and power boilers used in the pulp and paper industry. However, the program will also accommodate boilers common to other industries

Exact and approximate multiple diffraction calculations

International Nuclear Information System (INIS)

Alexander, Y.; Wallace, S.J.; Sparrow, D.A.

1976-08-01

A three-body potential scattering problem is solved in the fixed scatterer model exactly and approximately to test the validity of commonly used assumptions of multiple scattering calculations. The model problem involves two-body amplitudes that show diffraction-like differential scattering similar to high energy hadron-nucleon amplitudes. The exact fixed scatterer calculations are compared to Glauber approximation, eikonal-expansion results and a noneikonal approximation

Preliminary Hazard Classification for the 105-B Reactor

International Nuclear Information System (INIS)

Kerr, N.R.

1997-08-01

This document summarizes the inventories of radioactive and hazardous materials present within the 105-B Reactor and uses the inventory information to determine the preliminary hazard classification for the surveillance and maintenance activities of the facility. The result of this effort was the preliminary hazard classification for the 105-B Building surveillance and maintenance activities. The preliminary hazard classification was determined to be Nuclear Category 3. Additional hazard and accident analysis will be documented in a separate report to define the hazard controls and final hazard classification

Extensions to the SCDAP/RELAP5 code for the modeling of debris oxidation and materials interactions preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.; Davis, K.L.

1993-02-01

Preliminary designs are proposed for extending the SCDAP/RELAP5 code so that it models (a) the oxidation of slumping fuel rod material and cohesive and porous debris and (b) the interaction of PWR control rod materials with the other materials in a reactor core. These extensions have the purpose of improving the code's calculation of the damage progression and hydrogen production that takes place during the early phase of a severe accident

Calculation of Rydberg interaction potentials

International Nuclear Information System (INIS)

Weber, Sebastian; Büchler, Hans Peter; Tresp, Christoph; Urvoy, Alban; Hofferberth, Sebastian; Menke, Henri; Firstenberg, Ofer

2017-01-01

The strong interaction between individual Rydberg atoms provides a powerful tool exploited in an ever-growing range of applications in quantum information science, quantum simulation and ultracold chemistry. One hallmark of the Rydberg interaction is that both its strength and angular dependence can be fine-tuned with great flexibility by choosing appropriate Rydberg states and applying external electric and magnetic fields. More and more experiments are probing this interaction at short atomic distances or with such high precision that perturbative calculations as well as restrictions to the leading dipole–dipole interaction term are no longer sufficient. In this tutorial, we review all relevant aspects of the full calculation of Rydberg interaction potentials. We discuss the derivation of the interaction Hamiltonian from the electrostatic multipole expansion, numerical and analytical methods for calculating the required electric multipole moments and the inclusion of electromagnetic fields with arbitrary direction. We focus specifically on symmetry arguments and selection rules, which greatly reduce the size of the Hamiltonian matrix, enabling the direct diagonalization of the Hamiltonian up to higher multipole orders on a desktop computer. Finally, we present example calculations showing the relevance of the full interaction calculation to current experiments. Our software for calculating Rydberg potentials including all features discussed in this tutorial is available as open source. (tutorial)

Active cooling for downhole instrumentation: Preliminary analysis and system selection

Energy Technology Data Exchange (ETDEWEB)

Bennett, G.A.

1988-03-01

A feasibility study and a series of preliminary designs and analyses were done to identify candidate processes or cycles for use in active cooling systems for downhole electronic instruments. A matrix of energy types and their possible combinations was developed and the energy conversion process for each pari was identified. The feasibility study revealed conventional as well as unconventional processes and possible refrigerants and identified parameters needing further clarifications. A conceptual design or series od oesigns for each system was formulated and a preliminary analysis of each design was completed. The resulting coefficient of performance for each system was compared with the Carnot COP and all systems were ranked by decreasing COP. The system showing the best combination of COP, exchangeability to other operating conditions, failure mode, and system serviceability is chosen for use as a downhole refrigerator. 85 refs., 48 figs., 33 tabs.

Preliminary results on food consumption rates for off-site dose calculation of nuclear power plants

International Nuclear Information System (INIS)

Lee, Gab Bock; Chung, Yang Geun; Bang, Sun Young; Kang, Duk Won

2005-01-01

The Internal dose by food consumption mostly account for radiological dose of public around nuclear power plants(NPP). But, food consumption rate applied to off-site dose calculation in Korea which is the result of field investigation around Kori NPP by the KAERI in 1988. is not reflected of the latest dietary characteristics. The Ministry of Health and Welfare Affairs has investigated the food and nutrition of nations every 3 years based on the Law of National Health Improvement. To update the food consumption rates of the maximum individual, the analysis of the national food investigation results and field surveys around nuclear power plant sites have been carried out

Preliminary cost estimating for the nuclear industry

International Nuclear Information System (INIS)

Klumpar, I.V.; Soltz, K.M.

1985-01-01

The nuclear industry has higher costs for personnel, equipment, construction, and engineering than conventional industry, which means that cost estimation procedures may need adjustment. The authors account for the special technical and labor requirements of the nuclear industry in making adjustments to equipment and installation cost estimations. Using illustrative examples, they show that conventional methods of preliminary cost estimation are flexible enough for application to emerging industries if their cost structure is similar to that of the process industries. If not, modifications can provide enough engineering and cost data for a statistical analysis. 9 references, 14 figures, 4 tables

Space reactor preliminary mechanical design

International Nuclear Information System (INIS)

Meier, K.L.

1983-01-01

An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO 2 block core mounts, bolted collar fuel module restraints, and a BeO central plug

Closure and Sealing Design Calculation

International Nuclear Information System (INIS)

T. Lahnalampi; J. Case

2005-01-01

The purpose of the ''Closure and Sealing Design Calculation'' is to illustrate closure and sealing methods for sealing shafts, ramps, and identify boreholes that require sealing in order to limit the potential of water infiltration. In addition, this calculation will provide a description of the magma that can reduce the consequences of an igneous event intersecting the repository. This calculation will also include a listing of the project requirements related to closure and sealing. The scope of this calculation is to: summarize applicable project requirements and codes relating to backfilling nonemplacement openings, removal of uncommitted materials from the subsurface, installation of drip shields, and erecting monuments; compile an inventory of boreholes that are found in the area of the subsurface repository; describe the magma bulkhead feature and location; and include figures for the proposed shaft and ramp seals. The objective of this calculation is to: categorize the boreholes for sealing by depth and proximity to the subsurface repository; develop drawing figures which show the location and geometry for the magma bulkhead; include the shaft seal figures and a proposed construction sequence; and include the ramp seal figure and a proposed construction sequence. The intent of this closure and sealing calculation is to support the License Application by providing a description of the closure and sealing methods for the Safety Analysis Report. The closure and sealing calculation will also provide input for Post Closure Activities by describing the location of the magma bulkhead. This calculation is limited to describing the final configuration of the sealing and backfill systems for the underground area. The methods and procedures used to place the backfill and remove uncommitted materials (such as concrete) from the repository and detailed design of the magma bulkhead will be the subject of separate analyses or calculations. Post-closure monitoring will not

Preliminary Analysis and Selection of Mooring Solution Candidates

DEFF Research Database (Denmark)

Thomsen, Jonas Bjerg; Delaney, Martin

This report covers a preliminary analysis of mooring solutions candidates for four large floating wave energy converters. The work is part of the EUDP project “Mooring Solutions for Large Wave Energy Converters” and is the outcome of "Work Package 3: Preliminary Analysis". The report further...... compose the "Milestone 4: Report on results of preliminary analysis and selection of final candidates. The report is produced by Aalborg University with input from the partner WECs Floating Power Plant, KNSwing, LEANCON and Wave Dragon. Tension Technology International (TTI) has provided a significant...

Uncertainty calculations made easier

International Nuclear Information System (INIS)

Hogenbirk, A.

1994-07-01

The results are presented of a neutron cross section sensitivity/uncertainty analysis performed in a complicated 2D model of the NET shielding blanket design inside the ITER torus design, surrounded by the cryostat/biological shield as planned for ITER. The calculations were performed with a code system developed at ECN Petten, with which sensitivity/uncertainty calculations become relatively simple. In order to check the deterministic neutron transport calculations (performed with DORT), calculations were also performed with the Monte Carlo code MCNP. Care was taken to model the 2.0 cm wide gaps between two blanket segments, as the neutron flux behind the vacuum vessel is largely determined by neutrons streaming through these gaps. The resulting neutron flux spectra are in excellent agreement up to the end of the cryostat. It is noted, that at this position the attenuation of the neutron flux is about 1 l orders of magnitude. The uncertainty in the energy integrated flux at the beginning of the vacuum vessel and at the beginning of the cryostat was determined in the calculations. The uncertainty appears to be strongly dependent on the exact geometry: if the gaps are filled with stainless steel, the neutron spectrum changes strongly, which results in an uncertainty of 70% in the energy integrated flux at the beginning of the cryostat in the no-gap-geometry, compared to an uncertainty of only 5% in the gap-geometry. Therefore, it is essential to take into account the exact geometry in sensitivity/uncertainty calculations. Furthermore, this study shows that an improvement of the covariance data is urgently needed in order to obtain reliable estimates of the uncertainties in response parameters in neutron transport calculations. (orig./GL)

Development of fault parameters for use in risk assessment modeling in the Pasco Basin, Columbia Plateau, South Central Washington: a preliminary study

International Nuclear Information System (INIS)

Caggiano, J.A.

1982-03-01

Preliminary data on strain rate, seismicity, and estimated earthquake source parameters suggest limitations on the extent of postulated faulting and its impact on a nuclear waste repository in Columbia River basalt. Structural relief of dated basalt flows, attitude of Pliocene sediments, geodetic surveys, size and distribution of earthquakes, and focal mechanism solutions indicate that deformation of basalt under north-south compression was under way in the Miocene and has continued on existing structures at an average rate of much less than 1 mm/yr in the Pasco Basin. Lengths and displacements of mapped faults suggest limits on the postulated fault that could intersect a repository and produce an earthquake of about magnitude 6.5. Using a credible earthquake permits calculation of preliminary source parameters for risk assessment modeling during a single episode of slip on the postulated new fault and indicates displacement of less than or equal to 1, m on a steeply dipping fault of less than or equal to 50 km length could occur. Preliminary source parameter calculations suggest that displacements of less than or equal to 2 cm may occur during microearthquakes in swarms. The area of fault rupture may be tens of square meters up to a few square kilometers, suggesting slip on joints. Seismic moments for postulated earthquakes in the interconnecting fault and microearthquake scenarios compare favorably with reported values for similar-sized earthquakes in different media and suggest that the estimated fault parameters are reasonable until an adequate tectonic model has been developed

The Scope Calculation for the Distribution of the Plate-out in the 'OGL-1' experiment using MIDAS and Its model review

International Nuclear Information System (INIS)

Park, Jong-Hwa; Kim, Dong-Ha; Lee, Won-Jae

2007-01-01

The scope calculation on the plate-out in the HTGR related OGL-1 experiment using the MIDAS code was performed in the frame of the preliminary study to develop the MIDAS/GCR for simulating the plate-out, dust and tritium in a HTGR. From this scope calculation, the user specified type of the fission product vapor species in the circuit and the distribution of the circuit surface temperature were identified as the important factors that can have a strong effect on the distribution of fission product plate out over the HTGR loop. Also the analytical solution for calculating the plate-out by considering a radioactive decay was derived for MIDAS. These identified factors and the new analytical solution will be taken into account in developing the MIDAS/GCR

An adaptive sampling scheme for deep-penetration calculation

International Nuclear Information System (INIS)

Wang, Ruihong; Ji, Zhicheng; Pei, Lucheng

2013-01-01

As we know, the deep-penetration problem has been one of the important and difficult problems in shielding calculation with Monte Carlo Method for several decades. In this paper, an adaptive Monte Carlo method under the emission point as a sampling station for shielding calculation is investigated. The numerical results show that the adaptive method may improve the efficiency of the calculation of shielding and might overcome the under-estimation problem easy to happen in deep-penetration calculation in some degree

Towards a Quantum Dynamical Study of the H_2O+H_2O Inelastic Collision: Representation of the Potential and Preliminary Results

Science.gov (United States)

Ndengue, Steve Alexandre; Dawes, Richard

2017-06-01

Water, an essential ingredient of life, is prevalent in space and various media. H_2O in the gas phase is the major polyatomic species in the interstellar medium (ISM) and a primary target of current studies of collisional dynamics. In recent years a number of theoretical and experimental studies have been devoted to H_2O-X (with X=He, H_2, D_2, Ar, ?) elastic and inelastic collisions in an effort to understand rotational distributions of H_2O in molecular clouds. Although those studies treated several abundant species, no quantum mechanical calculation has been reported to date for a nonlinear polyatomic collider. We present in this talk the preliminary steps toward this goal, using the H_2O molecule itself as our collider, the very accurate MB-Pol surface to describe the intermolecular interaction and the MultiConfiguration Time Dependent (MCTDH) algorithm to study the dynamics. One main challenge in this effort is the need to express the Potential Energy Surface (PES) in a sum-of-products form optimal for MCTDH calculations. We will describe how this was done and present preliminary results of state-to-state probabilities.

Detection limit calculations for different total reflection techniques

International Nuclear Information System (INIS)

Sanchez, H.J.

2000-01-01

In this work, theoretical calculations of detection limits for different total-reflection techniques are presented.. Calculations include grazing incidence (TXRF) and gracing exit (GEXRF) conditions. These calculations are compared with detection limits obtained for conventional x-ray fluorescence (XRF). In order to compute detection limits the Shiraiwa and Fujino's model to calculate x-ray fluorescence intensities was used. This model made certain assumptions and approximations to achieve the calculations, specially in the case of the geometrical conditions of the sample, and the incident and takeoff beams. Nevertheless the calculated data of detection limits for conventional XRF and total-reflection XRF show a good agreement with previous results. The model proposed here allows to analyze the different sources of background and the influence of the excitation geometry, which contribute to the understanding of the physical processes involved in the XRF analysis by total reflection. Finally, a comparison between detection limits in total-reflection analysis at grazing incidence and at grazing exit is carried out. Here a good agreement with the theoretical predictions of the reversibility principle is found, showing that detection limits are similar for both techniques. (author)

A preliminary assessment of the radiological impact of the Chernobyl reactor accident on the population of the European Community

International Nuclear Information System (INIS)

Morrey, M.; Brown, J.; Williams, J.A.; Crick, M.J.; Simmonds, J.R.; Hill, M.D.

1988-01-01

Following the Chernobyl accident the Commission of the European Communities asked the National Radiological Protection Board to carry out a preliminary assessment of the radiological consequences of the accident on the population of the European Community (EC). The aim of the study was to review information on the environmental contamination measured in member states of the EC; to make a preliminary assessment of individual and population doses for each country; to make an estimate of the resulting health impact and to indicate the effects of the various countermeasures taken by member states in terms of the reductions in both individual and population exposure which they produced. All of the main pathways by which people have been and will be exposed to radiation as a result of the accident were included in the assessment. The impact estimate is based on environmental measurements made during the month after the accident, and on calculations made using mathematical models of radionuclide transfer through the environment. The calculated effective doses to average individuals in EC countries from exposure over the next 50 years range from 0.3 μSv (in Portugal) to between about 300 and 500 μSv (in the FRG, Italy and Greece). The total collective effective dose to the population of EC countries, integrated over all time, is estimated to be about 80 000 man Sv. This may be compared to the collective effective dose from natural background radiation of about 500 000 man Sv every year. In some countries, the restrictions placed on consumption of some foods are estimated to have been effective in reducing doses to the most exposed individuals; the reduction being up to about a factor of 2. The results presented in this paper should therefore be regarded as preliminary

Preliminary results of MR imaging of lymphoma: Distinguishing active tumor from benign residue

International Nuclear Information System (INIS)

Drace, J.; Baker, L.L.; Chang, P.; Castellino, R.A.

1987-01-01

Distinguishing tumor from benign posttreatment tissue based on both morphologic and tissue characteristics is critically important. Patients are studied before, during, and after treatment; at the time of recurrence; and on long-term follow-up. Multisection spin-echo sequences in orthogonal planes and a special single-section tissue characterization matrix of 16 different repetition time/echo time combinations are used. These basic images are used for cluster analysis (approximate fuzzy C means), T1-T2 synthetic images, linear combinations, and comparison with internal standards. Preliminary results in 35 patients imaged before treatment and 12 patients with follow-up examinations consistently show lymphoma masses to have complex architecture with high T2-weighted signal and moderate T1-weighted signal, distinct from posttreatment fibrosis. Uncommon components of active tumor with low T2-weighted signal appear distinct from fibrosis on T1-weighted images. Preliminary cluster analysis results show distinct clustering of active lymphoma versus fibrosis and biopsy-proved cystic degeneration

Core physics calculation and analysis for SNRE

International Nuclear Information System (INIS)

Xie Jiachun; Zhao Shouzhi; Jia Baoshan

2010-01-01

Five different precise calculation models have been set up for Small Nuclear Rocket Engine (SNRE) core based on MCNP code, and then the effective multiplication constant, drum control worth and power distribution were calculated. The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation, but a detailed description of elements have to be used in the element internal power distribution calculation. The results of physics parameters show that the basic characteristics of SNRE are reasonable. The drum control worth is sufficient. The power distribution is symmetrical and reasonable. All of the parameters can satisfy the design requirement. (authors)

Calculation of physical and thermo hydro-dynamic parameters of a thermal research reactor; Prorachun fizichkih i toplotno hidro-dinamichkih parametara termichkog istrazhivachkog reaktora

Energy Technology Data Exchange (ETDEWEB)

Matausek, M; Spasojevic, D; Jovic, V; Marinkovic, N [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

1988-07-01

The paper presents initial activities on creating a design concept of a new thermal research reactor, which should be built according to the research and development program in the field of nuclear fuel cycle technologies. For one possible type of such a reactor basic design parameters are specified and some preliminary results of nuclear, thermal and hydrodynamic design calculations are given. (author)

Ratio of thyroid radioiodine uptake calculated via the physic decay rate of the standard radioactive source: a preliminary study

International Nuclear Information System (INIS)

Zeng Yu; Zhou Luyi

2010-01-01

Objectives: To compare the difference of the ratio of thyroid radioiodine ( 131 I) uptake calculated by actually measuring counts of the standard radioactive source(method 1) and by computing counts of the standard radioactive source via physic half life of 131 I (method 2). Methods: Two hundred and nine consecutive patients with Graves' Disease were prospectively recruited. The ratio of thyroid 131 I uptake was calculated by two methods at 4 h and 24 h after administration of 1.48 MBq 131 I, respectively. Paired t-test was used to compare the difference between the two methods. Results: The ratio of thyroid 131 I uptake at 4h was (32±16)% and ( 35±10)% (t=1.98, P=0.20), at 24h (72±19)% and (69±24)% ( t=1.49, P=0.23), respectively, by the two methods. Conclusion: To calculate the ratio of thyroid 131 I uptake via the physic half life of the standard radioactive resource is feasible, and can both reduce the risk of ionizing radiation to technical staff and act as verifying method for quality control of thyroid function equipment. (authors)

Psychophysiological deficits in young adolescents with psychosis or ADHD: Preliminary findings

DEFF Research Database (Denmark)

Rydkjær, Jacob; Jepsen, Jens Richardt Møllegaard; Fagerlund, Birgitte

add valuable information on how to differentiate premature stages of early onset psychosis from ADHD. Aim: To characterize psychophysiological deficits in young adolescents with psychosis or ADHD and compare the profiles of impariments between the two groups. Materials and methods: A cohort of young...... and low intensity prepulse trials, Mismatch Negativity (MMN), Selective Attention (SA) and P50. Results: Preliminary analyses of 18 patients with psychosis and 12 patients with ADHD showed significantly less PPI in the higher intensity prepulse trials in the psychosis group than in the ADHD group....... No significant group difference was found in the lower intensity prepulse trials. Conclusion: The preliminary results indicate lower levels of PPI in adolescents with early onset psychosis than in young patients with ADHD. If these results hold in the final analyses then this knowledge may contribute to better...

The Monte Carlo applied for calculation dose

International Nuclear Information System (INIS)

Peixoto, J.E.

1988-01-01

The Monte Carlo method is showed for the calculation of absorbed dose. The trajectory of the photon is traced simulating sucessive interaction between the photon and the substance that consist the human body simulator. The energy deposition in each interaction of the simulator organ or tissue per photon is also calculated. (C.G.C.) [pt

Preliminary comparison of the system of AERMOD and ISCST3 models

International Nuclear Information System (INIS)

Turtos Carbonell, Leonor; Curbelo Garea, Lariza; Diaz Rivero, Norberto

2006-01-01

On October 21st, 2005 the U.S. Environmental Protection Agency (EPA), establishes AERMOD as regulatory model to be used for the dispersion of pollutants at local scale, in substitution of the ISCST3 model used up to that moment. Whenever a new dispersion model appears, it is necessary for the scientific community to make a comparison in order to discover the differences between the results obtained with the new model and the previous one. Considering the above mentioned fact, this work makes a preliminary comparison between the maximum concentrations calculated by each model (ISCST3 and AERMOD) for a specific case study that consists of eleven batteries of generation sets distributed throughout Havana City which will operate in base load mode and will use a fuel oil with 4% of sulphur. The modelling domain is the 50 xs 37 km with 1 x 1 km cells for a total of 1 850 calculation points (receptors), located in all Havana City and the bordering municipalities of Havana province. In each one of these receptors the dispersion of SO 2 and NO x were modelled

Optimizing Parameters of Axial Pressure-Compounded Ultra-Low Power Impulse Turbines at Preliminary Design

Science.gov (United States)

Kalabukhov, D. S.; Radko, V. M.; Grigoriev, V. A.

2018-01-01

Ultra-low power turbine drives are used as energy sources in auxiliary power systems, energy units, terrestrial, marine, air and space transport within the confines of shaft power N td = 0.01…10 kW. In this paper we propose a new approach to the development of surrogate models for evaluating the integrated efficiency of multistage ultra-low power impulse turbine with pressure stages. This method is based on the use of existing mathematical models of ultra-low power turbine stage efficiency and mass. It has been used in a method for selecting the rational parameters of two-stage axial ultra-low power turbine. The article describes the basic features of an algorithm for two-stage turbine parameters optimization and for efficiency criteria evaluating. Pledged mathematical models are intended for use at the preliminary design of turbine drive. The optimization method was tested at preliminary design of an air starter turbine. Validation was carried out by comparing the results of optimization calculations and numerical gas-dynamic simulation in the Ansys CFX package. The results indicate a sufficient accuracy of used surrogate models for axial two-stage turbine parameters selection

BioXTAS RAW, a software program for high-throughput automated small-angle X-ray scattering data reduction and preliminary analysis

DEFF Research Database (Denmark)

Nielsen, S.S.; Toft, K.N.; Snakenborg, Detlef

2009-01-01

A fully open source software program for automated two-dimensional and one-dimensional data reduction and preliminary analysis of isotropic small-angle X-ray scattering (SAXS) data is presented. The program is freely distributed, following the open-source philosophy, and does not rely on any...... commercial software packages. BioXTAS RAW is a fully automated program that, via an online feature, reads raw two-dimensional SAXS detector output files and processes and plots data as the data files are created during measurement sessions. The software handles all steps in the data reduction. This includes...... mask creation, radial averaging, error bar calculation, artifact removal, normalization and q calibration. Further data reduction such as background subtraction and absolute intensity scaling is fast and easy via the graphical user interface. BioXTAS RAW also provides preliminary analysis of one...

Preliminary safety evaluation, based on initial site investigation data. Planning document

International Nuclear Information System (INIS)

Hedin, Allan

2002-12-01

This report is a planning document for the preliminary safety evaluations (PSE) to be carried out at the end of the initial stage of SKBs ongoing site investigations for a deep repository for spent nuclear fuel. The main purposes of the evaluations are to determine whether earlier judgements of the suitability of the candidate area for a deep repository with respect to long-term safety holds up in the light of borehole data and to provide feed-back to continued site investigations and site specific repository design. The preliminary safety evaluations will be carried out by a safety assessment group, based on a site model, being part of a site description, provided by a site modelling group and a repository layout within that model suggested by a repository engineering group. The site model contains the geometric features of the site as well as properties of the host rock. Several alternative interpretations of the site data will likely be suggested. Also the biosphere is included in the site model. A first task for the PSE will be to compare the rock properties described in the site model to previously established criteria for a suitable host rock. This report gives an example of such a comparison. In order to provide more detailed feedback, a number of thermal, hydrological, mechanical and chemical analyses of the site will also be included in the evaluation. The selection of analyses is derived from the set of geosphere and biosphere analyses preliminarily planned for the comprehensive safety assessment named SR-SITE, which will be based on a complete site investigation. The selection is dictated primarily by the expected feedback to continued site investigations and by the availability of data after the PSE. The repository engineering group will consider several safety related factors in suggesting a repository layout: Thermal calculations will be made to determine a minimum distance between canisters avoiding canister surface temperatures above 100 deg C

TREE STEM RECONSTRUCTION USING VERTICAL FISHEYE IMAGES: A PRELIMINARY STUDY

Directory of Open Access Journals (Sweden)

A. Berveglieri

2016-06-01

Full Text Available A preliminary study was conducted to assess a tree stem reconstruction technique with panoramic images taken with fisheye lenses. The concept is similar to the Structure from Motion (SfM technique, but the acquisition and data preparation rely on fisheye cameras to generate a vertical image sequence with height variations of the camera station. Each vertical image is rectified to four vertical planes, producing horizontal lateral views. The stems in the lateral view are rectified to the same scale in the image sequence to facilitate image matching. Using bundle adjustment, the stems are reconstructed, enabling later measurement and extraction of several attributes. The 3D reconstruction was performed with the proposed technique and compared with SfM. The preliminary results showed that the stems were correctly reconstructed by using the lateral virtual images generated from the vertical fisheye images and with the advantage of using fewer images and taken from one single station.

RELAP5/MOD2 blind calculation of GERDA small break test and data comparison

International Nuclear Information System (INIS)

Ogden, D.M.; Steiner, J.L.; Waterman, M.E.

1985-01-01

The Idaho National Engineering Laboratory (INEL), in support of the USNRC, has developed a RELAP5/MOD2 model of the GERDA facility to be used for analysis of the GERDA data, particularly relative to the phenomena of natural circulation and the boiler condenser mode of heat transfer. A blind calculation of GERDA Test 1605AA and a preliminary comparison with experimental data has been performed. The GERDA facility is a single loop integral facility with an electrically heated core. A general arrangement diagram of the facility is shown. The GERDA facility was designed for the performance of both separate effects and overall systems tests

Stress analysis of HLW containers. Preliminary ring test exercise Compas project

International Nuclear Information System (INIS)

1989-01-01

This document describes the series of experiments and associated calculations performed as the Compas preliminary ring test exercise. A number of mild steel rings, representative of sections through HLW containers, some notched and pre-cracked, were tested in compression right up to and beyond their ultimate load. The Compas project partners independently modelled the behaviour of these rings using their finite element codes. Four different ring types were tested, and each test was repeated three times. For three of the ring types, the three test repetitions gave identical results. The fourth ring, which was not modelled by the partners, had a 4 mm thick layer of weld metal deposited on its surface. The three tests on this ring did not give identical results and suggested that the effect of welding methods should be addressed at a later stage of the project. Fracture was not found to be a significant cause of ring failure. The results of the ring tests were compared with the partners predictions, and additionally some time was spent assessing where the use of the codes could be improved. This exercise showed that the partners codes have the ability to produce results within acceptable limits. Most codes were unable to model stable crack growth. There were indications that some codes would not be able to cope with a significantly more complex three-dimensional analysis

Preliminary design report for the NAC combined transport cask

International Nuclear Information System (INIS)

1990-04-01

Nuclear Assurance Corporation (NAC) is under contract to the United States Department of Energy (DOE) to design, license, develop and test models, and fabricate a prototype cask transportation system for nuclear spent fuel. The design of this combined transport (rail/barge) transportation system has been divided into two phases, a preliminary design phase and a final design phase. This Preliminary Design Package (PDP) describes the NAC Combined Transport Cask (NAC-CTC), the results of work completed during the preliminary design phase and identifies the additional detailed analyses, which will be performed during final design. Preliminary analytical results are presented in the appropriate sections and supplemented by summaries of procedures and assumptions for performing the additional detailed analyses of the final design. 60 refs., 1 fig., 2 tabs

Preliminary study of magnet design for an SSC

International Nuclear Information System (INIS)

Taylor, C.E.; Meuser, R.B.

1983-08-01

The overriding design consideration for the SSC magnets is that cost of the facility be minimized; at 8 T, approximately 40 km of bending magnets is required for each ring of a 20 TeV collider. We present some results of a parametric study of two-in-one, iron-core magnets for an SSC. These results are necessarily preliminary in nature, and are intended only to show some of the trade-offs for a wide range of the variables. We show also some results for a reference design that produces 6.5 T in the aperture at 4.4 K for a coil inside diameter of 40 mm. It is not to be inferred that we have established this to be an optimum in any sense

45 CFR 671.15 - Publication of preliminary determination

Science.gov (United States)

2010-10-01

... 45 Public Welfare 3 2010-10-01 2010-10-01 false Publication of preliminary determination 671.15... Publication of preliminary determination Prior to any designation or redesignation of substances pursuant to... Environmental Protection Agency and other federal agencies, within 30 days after the date of publication of...

Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

Energy Technology Data Exchange (ETDEWEB)

Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwai, Takehiko

1998-07-01

A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

Stokes polarimetry probe for skin lesion evaluation: preliminary results

Science.gov (United States)

Louie, Daniel C.; Tchvialeva, Lioudmilla; Kalia, Sunil; Lui, Harvey; Lee, Tim K.

2018-02-01

This paper reports on the design of a prototype in-vivo Stokes polarimetry probe for skin lesion evaluation, and preliminary results from skin phantom and clinical trials of this device. The probe releases a single millisecond-long pulse from a laser diode with either linear or circular polarization. It then captures the resulting backscattered far-field polarization speckle and calculates the Stokes parameters. This probe was designed with three novel innovations in mind. First, the Stokes vector is captured quickly, using low-cost components without the use of moving parts. Second, a compact collimated laser diode was used as the light source. Third, the device and detector geometry were designed to produce and capture a uniform speckle field. In the first clinical trial of this device, measurements were taken from a variety of skin lesions, both cancerous and benign. The Stokes vector was measured and used to calculate the degree of polarization (DOP), the azimuth angle, and the ellipticity angle of the polarization ellipse for two input light polarizations. Among other findings, the DOP for circular polarized input light was consistently lower than the DOP for linear polarized input light. These findings indicate the potential for a fast and low-cost in-vivo skin cancer screening tool, and encourages the continuing development of this probe's techniques.

Calculation of positron characteristics for elements of the periodic table

International Nuclear Information System (INIS)

Campillo Robles, J M; Ogando, E; Plazaola, F

2011-01-01

Positron characteristics have been calculated in bulk and monovacancies for most of the elements of the periodic table. Self-consistent and non-self-consistent schemes have been used for the calculation of the electronic structure in the solid, and different parametrizations for the positron enhancement factor and correlation energy. As it is known, positron lifetimes in bulk show a periodic behaviour with atomic number. These calculations also confirm that monovacancy lifetimes follow the same behaviour. The results obtained have been compared with selected experimental lifetime data, which confirms the calculated theoretical trends. Positron binding energies to a monovacancy have been calculated also for most of the elements of the periodic table. The binding energy shows a periodic behaviour with atomic number too.

78 FR 64478 - Request for Comments on the Preliminary Cybersecurity Framework

Science.gov (United States)

2013-10-29

...-01] Request for Comments on the Preliminary Cybersecurity Framework AGENCY: National Institute of... Cybersecurity Framework (``preliminary Framework''). The preliminary Framework was developed by NIST using... Cybersecurity'' (``Executive Order''). Under the Executive Order, the Secretary of Commerce is tasked to direct...

Preliminary psycometric assessment of the Brazilian version of the DISABKIDS Atopic Dermatitis Module.

Science.gov (United States)

Deon, Keila Cristiane; Santos, Danielle Maria de Souza Sério dos; Bullinger, Monika; Santos, Claudia Benedita dos

2011-12-01

To assess preliminary psychometric properties of the Brazilian Portuguese version of a questionnaire for measuring health-related quality of life in children and adolescents with atopic dermatitis. Cross-sectional study with a sample consisting of 52 children and adolescents aged 8 to 18 diagnosed with atopic dermatitis, and their parents or caregivers, selected at the dermatology department of a university hospital in the city of São Paulo, Southeast Brazil, in 2009. Construct validity, internal consistency and agreement between the responses of children and adolescents and their parents or caregivers were assessed in the Brazilian Portuguese version of the DISABKIDS-Atopic Dermatitis Module (ADM). Adequate internal consistency was found with Cronbach's alpha coefficients of 0.7024/0.8124 and 0.7239/0.8604. The multitrait multimethod analysis for assessing convergent validity showed measures higher than 0.30 for all items. The analysis showed good discriminant validity. Agreement between child self-report and parent proxy-report was evaluated using intra-class correlation with measures impact and social stigma of disease of 0.8173 and 0.7629, respectively. The study results showed that the DISABKIDS-ADM can be used by Brazilian researchers after its complete validation as it showed adequate preliminary psychometric properties and can be considered a valid, reliable instrument.

Preliminary Analysis of the Bundle-Duct Interaction for the fuel of SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

BDI (Bundle-Duct Interaction) occurs in the fuel of SFR (Sodium-cooled Fast Reactor) due to the radial expansion and bowing of a fuel pin bundle. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore, BDI, which is the dominant deformation mechanisms in a fuel pin bundle, should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE and BMBOO, have been developed to evaluate the BDI behavior. The bundle duct interaction model is also being developed for SFR in Korea. This model is based on ANSYS. In this paper, the fuel pin configuration model for the BDI calculation was established. The preliminary analysis of the bundle-duct interaction was performed to evaluate the fuel design concept.

A knowledge-based design framework for airplane conceptual and preliminary design

Science.gov (United States)

Anemaat, Wilhelmus A. J.

The goal of work described herein is to develop the second generation of Advanced Aircraft Analysis (AAA) into an object-oriented structure which can be used in different environments. One such environment is the third generation of AAA with its own user interface, the other environment with the same AAA methods (i.e. the knowledge) is the AAA-AML program. AAA-AML automates the initial airplane design process using current AAA methods in combination with AMRaven methodologies for dependency tracking and knowledge management, using the TechnoSoft Adaptive Modeling Language (AML). This will lead to the following benefits: (1) Reduced design time: computer aided design methods can reduce design and development time and replace tedious hand calculations. (2) Better product through improved design: more alternative designs can be evaluated in the same time span, which can lead to improved quality. (3) Reduced design cost: due to less training and less calculation errors substantial savings in design time and related cost can be obtained. (4) Improved Efficiency: the design engineer can avoid technically correct but irrelevant calculations on incomplete or out of sync information, particularly if the process enables robust geometry earlier. Although numerous advancements in knowledge based design have been developed for detailed design, currently no such integrated knowledge based conceptual and preliminary airplane design system exists. The third generation AAA methods are tested over a ten year period on many different airplane designs. Using AAA methods will demonstrate significant time savings. The AAA-AML system will be exercised and tested using 27 existing airplanes ranging from single engine propeller, business jets, airliners, UAV's to fighters. Data for the varied sizing methods will be compared with AAA results, to validate these methods. One new design, a Light Sport Aircraft (LSA), will be developed as an exercise to use the tool for designing a new airplane

Parallel computational in nuclear group constant calculation

International Nuclear Information System (INIS)

Su'ud, Zaki; Rustandi, Yaddi K.; Kurniadi, Rizal

2002-01-01

In this paper parallel computational method in nuclear group constant calculation using collision probability method will be discuss. The main focus is on the calculation of collision matrix which need large amount of computational time. The geometry treated here is concentric cylinder. The calculation of collision probability matrix is carried out using semi analytic method using Beckley Naylor Function. To accelerate computation speed some computer parallel used to solve the problem. We used LINUX based parallelization using PVM software with C or fortran language. While in windows based we used socket programming using DELPHI or C builder. The calculation results shows the important of optimal weight for each processor in case there area many type of processor speed

UVISS preliminary visibility analysis

DEFF Research Database (Denmark)

Betto, Maurizio

1998-01-01

The goal of this work is to obtain a preliminary assessment of the sky visibility for anastronomical telescope located on the express pallet of the International SpaceStation (ISS)} taking into account the major constraints imposed on the instrument by the ISSattitude and structure. Part of the w......The goal of this work is to obtain a preliminary assessment of the sky visibility for anastronomical telescope located on the express pallet of the International SpaceStation (ISS)} taking into account the major constraints imposed on the instrument by the ISSattitude and structure. Part...... of the work is also to setup the kernel of a software tool for the visibility analysis thatshould be easily expandable to consider more complex strucures for future activities.This analysis is part of the UVISS assessment study and it is meant to provide elementsfor the definition and the selection...

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.

Process heat applications of HTR-PM600 in Chinese petrochemical industry: Preliminary study of adaptability and economy

International Nuclear Information System (INIS)

Fang, Chao; Min, Qi; Yang, Yanran; Sun, Yuliang

2017-01-01

Highlights: •High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. •The joint of a 600 MW modular HTGR (HTR-PM600) and petrochemical industry is achievable. •The mature technology of turbine in thermal power station could be readily adopted. •The economy of this scheme is also acceptable. -- Abstract: High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. In this article, the preliminary feasibility of a 600 MW modular HTGR (HTR-PM600) working as heat source for a typical hypothetical Chinese petrochemical factory is discussed and it is found that the joint of HTR-PM600 and petrochemical industry is achievable. In detail, the heat and water balance analysis of the petrochemical factory is given. Furthermore, the direct cost of heat supplied by HTR-PM600 is calculated and corresponding economy is estimated. The results show that though there are several challenges, the application of process heat of HTGR to petrochemical industry is practical in sense of both technology and economy.

Molecular calculations with B functions

International Nuclear Information System (INIS)

Steinborn, E.O.; Homeier, H.H.H.; Ema, I.; Lopez, R.; Ramirez, G.

2000-01-01

A program for molecular calculations with B functions is reported and its performance is analyzed. All the one- and two-center integrals and the three-center nuclear attraction integrals are computed by direct procedures, using previously developed algorithms. The three- and four-center electron repulsion integrals are computed by means of Gaussian expansions of the B functions. A new procedure for obtaining these expansions is also reported. Some results on full molecular calculations are included to show the capabilities of the program and the quality of the B functions to represent the electronic functions in molecules

Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

2016-11-01

Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

Gas cooled fast reactor 2400 MWTh, status on the conceptual design studies and preliminary safety analysis

International Nuclear Information System (INIS)

Malo, J.Y.; Alpy, N.; Bentivoglio, F.

2009-01-01

The Gas cooled Fast Reactor (GFR) is considered by the French Commissariat a l'Energie Atomique as a promising concept, combining the benefits of fast spectrum and high temperature, using Helium as coolant. A status on the GFR preliminary viability was made at the end of 2007, ending the pre-conceptual design phase. A consistent overall systems arrangement was proposed and a preliminary safety analysis based on operating transient calculations and a simplified PSA had established a global confidence in the feasibility and safety of this baseline concept. Its potential for attractive performances had been pointed out. Compare to the more mature Sodium Fast Reactor technology, no demonstrator has ever been built and the feasibility demonstration will required a longer lead time. The next main project milestone is related to the GFR viability, scheduled in 2012. The current studies consist in revisiting the reactor reference design options as selected at the end of 2007. Most of them are being consolidated by going more in depth in the analysis. Some possible alternatives are assessed. The paper will give a status on the last studies performed on the core design and corresponding neutronics and cycle performance, the Decay Heat Removal strategy and preliminary safety analysis, systems design and balance of plant... This paper is complementary to the Icapp'09 papers 9062 dealing with the Gas cooled Fast Reactor Demonstrator ALLEGRO and 9378 related to GFR transients analysis. (author)

National Data Center Preparedness Exercise 2015 (NPE 2015): MY-NDC Preliminary Analysis Result

International Nuclear Information System (INIS)

Faisal Izwan Abdul Rashid; Muhammed Zulfakar Zolkaffly

2016-01-01

Malaysia has established the CTBT National Data Centre (MY-NDC) in December 2005. MY-NDC is tasked to perform Comprehensive Nuclear-Test-Ban-Treaty (CTBT) data management as well as provide information for Treaty related events to Nuclear Malaysia as CTBT National Authority. In 2015, MY-NDC has participated in the National Data Centre Preparedness Exercise 2015 (NPE 2015). This paper aims at presenting MY-NDC preliminary analysis result of NPE 2015. In NPE 2015, MY-NDC has performed five different analyses, namely, radionuclide, atmospheric transport modelling (ATM), data fusion, seismic analysis and site forensics. The preliminary findings show the hypothetical scenario in NPE 2015 most probably is an uncontained event resulted high release of radionuclide to the air. (author)

Preliminary Coupling of MATRA Code for Multi-physics Analysis

International Nuclear Information System (INIS)

Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

2014-01-01

The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

Preliminary risk assessment for nuclear waste disposal in space. Volume I. Executive summary of technical report

International Nuclear Information System (INIS)

Rice, E.E.; Denning, R.S.; Friedlander, A.L.

1982-01-01

Three major conclusions come from this preliminary risk assessment of nuclear waste disposal in space. Preliminary estimates of space disposal risk are low, even with the estimated uncertainty bounds. If calculated mined geologic repository (MGR) release risks remain low, and the EPA requirements continue to be met, then no additional space disposal study effort is warranted. If risks perceived by the public are significant in the acceptance of mined geologic repositories, then consideration of space disposal as an MGR complement is warranted. As a result of this study, the following recommendations are made to NASA and the US DOE: During the continued evaluation of the mined geologic repository risk over the years ahead by DOE, if any significant increase in the calculated health risk is predicted for the MGR, then space disposal should be reevaluated at that time. The risks perceived by the public for the MGR should be evaluated on a broad basis by an independent organization to evaluate acceptance. If, in the future, MGR risks are found to be significant due to some presently unknown technical or social factor, and space disposal is selected as an alternative that may be useful in mitigating the risks, then the following space disposal study activities are recommended: improvement in chemical processing technology for wastes; payload accident response analysis; risk uncertainty analysis for both MGR and space disposal; health risk modeling that includes pathway and dose estimates; space disposal cost modeling; assessment of space disposal perceived (by public) risk benefit; and space systems analysis supporting risk and cost modeling

Preliminary results on the Arecibo Pisces-Perseus Supercluster Survey

Science.gov (United States)

Cortes, Rosemary; Lebron, Mayra; Jones, Michael G.; Koopmann, Rebecca A.; Haynes, Martha P.; APPSS Team, Undergraduate ALFALFA Team, and the ALFALFA Team

2018-01-01

The Arecibo Pisces-Perseus Supercluster Survey (APPSS) aims to exploit the Baryonic Tully-Fisher Relation to derive distances and peculiar velocities of galaxies in and near the main ridge of the Pisces-Perseus Supercluster (PPS), one of the most prominent features of the Cosmic Web in the nearby Universe. The sample of galaxies contains ~ 600 sources in the low-mass range (8 Team institutions in which each group contributes to the analysis of a subset of the HI PPS data. In this poster, we will present the contributions of the U.P.R. team to the APPSS project. We will show the procedure used for the Arecibo HI data analysis, including some examples, and will show our preliminary results.

Preliminary evaluation of uranium deposits. A geostatistical study of drilling density in Wyoming solution fronts

International Nuclear Information System (INIS)

Sandefur, R.L.; Grant, D.C.

1976-01-01

Studies of a roll-front uranium deposit in Shirley Basin Wyoming indicate that preliminary evaluation of the reserve potential of an ore body is possible with less drilling than currently practiced in industry. Estimating ore reserves from sparse drilling is difficult because most reserve calculation techniques do not give the accuracy of the estimate. A study of several deposits with a variety of drilling densities shows that geostatistics consistently provides a method of assessing the accuracy of an ore reserve estimate. Geostatistics provides the geologist with an additional descriptive technique - one which is valuable in the economic assessment of a uranium deposit. Closely spaced drilling on past properties provides both geological and geometric insight into the occurrence of uranium in roll-front type deposits. Just as the geological insight assists in locating new ore bodies and siting preferential drill locations, the geometric insight can be applied mathematically to evaluate the accuracy of a new ore reserve estimate. By expressing the geometry in numerical terms, geostatistics extracts important geological characteristics and uses this information to aid in describing the unknown characteristics of a property. (author)

Graphical comparison of calculated internal conversion coefficients

International Nuclear Information System (INIS)

Ewbank, W.B.

1980-11-01

Calculated values of the coefficients of internal conversion of gamma rays in the K shell and L 1 , L 2 , L 3 subshells from published tabulations by Band and Trzhaskovskaya and by Roesel et al. at Data Nucl. Data Tables, 21, 92-514(1978) are compared with values obtained by computer interpolation among tabulated values of Hager and Seltzer Nucl. Data, A4, 1-235(1968). In some cases, agreement among the three calculations is remarkably good, and differences are generally less than 5%. In a few cases, there are differences as large as 20 to 50%, corresponding to the threshold effect described by Roesel et al. The Z-dependent resonance minimum described by Roesel et al. is also observed in the comparison of E1-E4 conversion in the L 1 subshell. In several cases (notably M1-M4 conversion in the K shell and L 1 subshell), the Band and Roesel calculations show dramatically different dependence on gamma energy and atomic number. For Z = 100, the Band calculation for E4 conversion in the L 3 subshell shows irregular behavior at energies below the K-shell binding energy. A few high-quality measurements of internal conversion coefficients (+-5%) would help greatly to establish a basis for choice among the theoretical calculations. 32 figures

Emotion Knowledge and Attentional Differences in Preschoolers Showing Context-Inappropriate Anger.

Science.gov (United States)

Locke, Robin L; Lang, Nichole J

2016-08-01

Some children show anger inappropriate for the situation based on the predominant incentives, which is called context-inappropriate anger. Children need to attend to and interpret situational incentives for appropriate emotional responses. We examined associations of context-inappropriate anger with emotion recognition and attention problems in 43 preschoolers (42% male; M age = 55.1 months, SD = 4.1). Parents rated context-inappropriate anger across situations. Teachers rated attention problems using the Child Behavior Checklist-Teacher Report Form. Emotion recognition was ability to recognize emotional faces using the Emotion Matching Test. Anger perception bias was indicated by anger to non-anger situations using an adapted Affect Knowledge Test. 28% of children showed context-inappropriate anger, which correlated with lower emotion recognition (β = -.28) and higher attention problems (β = .36). Higher attention problems correlated with more anger perception bias (β = .32). This cross-sectional, correlational study provides preliminary findings that children with context-inappropriate anger showed more attention problems, which suggests that both "problems" tend to covary and associate with deficits or biases in emotion knowledge. © The Author(s) 2016.

A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

Calculated apparent yields of rare gas fission products

International Nuclear Information System (INIS)

Delucchi, A.A.

1975-01-01

The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)

ROUND-ROBIN ATOM-PROBE EXPERIMENT : PRELIMINARY RESULTS IN JAPAN

OpenAIRE

Nakamura , S.

1986-01-01

A round-robin experiment were to be carried out by 6 laboratories in Japan (Nishikawa ; Tokyo Inst. Tech., Sakurai and Igata ; Univ. of Tokyo, Ishikawa ; Hitachi, Tanino ; Nippon Steel Corp. and Nakamura ; Osaka Univ.) under the normal operating condition (T < l00 °K,. pulse fraction ~ 15% ~, P < 10-9 torr). Fe-Cr-Al and W-25%Re alloys, which are divided from a single wire were chosen as the specimen materials. A preliminary analysis of the W-Re alloy of the laboratory concerned show the good...

Radionuclide inventory calculation in VVER and BWR reactor

International Nuclear Information System (INIS)

Bouhaddane, A.; Farkas, F.; Slugen, V.; Ackermann, L.; Schienbein, M.

2014-01-01

The paper shows different aspects in the radionuclide inventory determination. Precise determination of the neutron flux distribution, presented for a BRW reactor, is vital for the activation calculations. The precision can be improved utilizing variance reduction methods as importance treatment, weight windows etc. Direct calculation of the radionuclide inventory via Monte Carlo code is presented for a VVER reactor. Burn-up option utilized in this calculation appears to be proper for reactor internal components. However, it will not be probably effective outside the reactor core. Further calculations in this area are required to support the forth-set findings. (authors)

Preliminary remediation goals for use at the U.S. Department of Energy Oak Ridge Operations Office

International Nuclear Information System (INIS)

1995-06-01

This report presents Preliminary Remediation Goals (PRGs) for use in human health risk assessment efforts under the United States Department of Energy, Oak Ridge Operations Office Environmental Restoration (ER) Division. Chemical-specific PRGs are concentration goals for individual chemicals for specific medium and land use combinations. The PRGs are referred to as risk-based because they have been calculated using risk assessment procedures. Risk-based calculations set concentration limits using both carcinogenic or noncarcinogenic toxicity values under specific exposure pathways. The PRG is a concentration that is derived from a specified excess cancer risk level or hazard quotient. This report provides the ER Division with standardized PRGs which are integral to the Remedial Investigation/Feasibility Study process. By managing the assumptions and systems used in PRG derivation, the Environmental Restoration Risk Assessment Program will be able to control the level of quality assurance associated with these risk-based guideline values

Preliminary remediation goals for use at the U.S. Department of Energy Oak Ridge Operations Office

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

This report presents Preliminary Remediation Goals (PRGs) for use in human health risk assessment efforts under the United States Department of Energy, Oak Ridge Operations Office Environmental Restoration (ER) Division. Chemical-specific PRGs are concentration goals for individual chemicals for specific medium and land use combinations. The PRGs are referred to as risk-based because they have been calculated using risk assessment procedures. Risk-based calculations set concentration limits using both carcinogenic or noncarcinogenic toxicity values under specific exposure pathways. The PRG is a concentration that is derived from a specified excess cancer risk level or hazard quotient. This report provides the ER Division with standardized PRGs which are integral to the Remedial Investigation/Feasibility Study process. By managing the assumptions and systems used in PRG derivation, the Environmental Restoration Risk Assessment Program will be able to control the level of quality assurance associated with these risk-based guideline values.

SU-F-T-111: Investigation of the Attila Deterministic Solver as a Supplement to Monte Carlo for Calculating Out-Of-Field Radiotherapy Dose

Energy Technology Data Exchange (ETDEWEB)

Mille, M; Lee, C [Division of Cancer Epidemiology and Genetics, National Cancer Institute, National Institutes of Health, Rockville, MD (United States); Failla, G [Varian Medical Systems, Gig Harbor, WA (United States)

2016-06-15

Purpose: To use the Attila deterministic solver as a supplement to Monte Carlo for calculating out-of-field organ dose in support of epidemiological studies looking at the risks of second cancers. Supplemental dosimetry tools are needed to speed up dose calculations for studies involving large-scale patient cohorts. Methods: Attila is a multi-group discrete ordinates code which can solve the 3D photon-electron coupled linear Boltzmann radiation transport equation on a finite-element mesh. Dose is computed by multiplying the calculated particle flux in each mesh element by a medium-specific energy deposition cross-section. The out-of-field dosimetry capability of Attila is investigated by comparing average organ dose to that which is calculated by Monte Carlo simulation. The test scenario consists of a 6 MV external beam treatment of a female patient with a tumor in the left breast. The patient is simulated by a whole-body adult reference female computational phantom. Monte Carlo simulations were performed using MCNP6 and XVMC. Attila can export a tetrahedral mesh for MCNP6, allowing for a direct comparison between the two codes. The Attila and Monte Carlo methods were also compared in terms of calculation speed and complexity of simulation setup. A key perquisite for this work was the modeling of a Varian Clinac 2100 linear accelerator. Results: The solid mesh of the torso part of the adult female phantom for the Attila calculation was prepared using the CAD software SpaceClaim. Preliminary calculations suggest that Attila is a user-friendly software which shows great promise for our intended application. Computational performance is related to the number of tetrahedral elements included in the Attila calculation. Conclusion: Attila is being explored as a supplement to the conventional Monte Carlo radiation transport approach for performing retrospective patient dosimetry. The goal is for the dosimetry to be sufficiently accurate for use in retrospective

Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

International Nuclear Information System (INIS)

Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

2003-01-01

The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

The development and validation of control rod calculation methods

International Nuclear Information System (INIS)

Rowlands, J.L.; Sweet, D.W.; Franklin, B.M.

1979-01-01

Fission rate distributions have been measured in the zero power critical facility, ZEBRA, for a series of eight different arrays of boron carbide control rods. Diffusion theory calculations have been compared with these measurements. The normalised fission rates differ by up to about 30% in some regions, between the different arrays, and these differences are well predicted by the calculations. A development has been made to a method used to produce homogenised cross sections for lattice regions containing control rods. Calculations show that the method also reproduces the reaction rate within the rod and the fission rate dip at the surface of the rod in satisfactory agreement with the more accurate calculations which represent the fine structure of the rod. A comparison between diffusion theory and transport theory calculations of control rod reactivity worths in the CDFR shows that for the standard design method the finite mesh approximation and the difference between diffusion theory and transport theory (the transport correction) tend to cancel and result in corrections to be applied to the standard mesh diffusion theory calculations of about +- 2% or less. This result applies for mesh centred finite difference diffusion theory codes and for the arrays of natural boron carbide control rods for which the calculations were made. Improvements have also been made to the effective diffusion coefficients used in diffusion theory calculations for control rod followers and these give satisfactory agreement with transport theory calculations. (U.K.)

Importance iteration in MORSE Monte Carlo calculations

International Nuclear Information System (INIS)

Kloosterman, J.L.; Hoogenboom, J.E.

1994-01-01

An expression to calculate point values (the expected detector response of a particle emerging from a collision or the source) is derived and implemented in the MORSE-SGC/S Monte Carlo code. It is outlined how these point values can be smoothed as a function of energy and as a function of the optical thickness between the detector and the source. The smoothed point values are subsequently used to calculate the biasing parameters of the Monte Carlo runs to follow. The method is illustrated by an example that shows that the obtained biasing parameters lead to a more efficient Monte Carlo calculation

Importance iteration in MORSE Monte Carlo calculations

International Nuclear Information System (INIS)

Kloosterman, J.L.; Hoogenboom, J.E.

1994-02-01

An expression to calculate point values (the expected detector response of a particle emerging from a collision or the source) is derived and implemented in the MORSE-SGC/S Monte Carlo code. It is outlined how these point values can be smoothed as a function of energy and as a function of the optical thickness between the detector and the source. The smoothed point values are subsequently used to calculate the biasing parameters of the Monte Carlo runs to follow. The method is illustrated by an example, which shows that the obtained biasing parameters lead to a more efficient Monte Carlo calculation. (orig.)

Analysis of Radiation Treatment Planning by Dose Calculation and Optimization Algorithm

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Sup; Yoon, In Ha; Lee, Woo Seok; Baek, Geum Mun [Dept. of Radiation Oncology, Asan Medical Center, Seoul (Korea, Republic of)

2012-09-15

Analyze the Effectiveness of Radiation Treatment Planning by dose calculation and optimization algorithm, apply consideration of actual treatment planning, and then suggest the best way to treatment planning protocol. The treatment planning system use Eclipse 10.0. (Varian, USA). PBC (Pencil Beam Convolution) and AAA (Anisotropic Analytical Algorithm) Apply to Dose calculation, DVO (Dose Volume Optimizer 10.0.28) used for optimized algorithm of Intensity Modulated Radiation Therapy (IMRT), PRO II (Progressive Resolution Optimizer V 8.9.17) and PRO III (Progressive Resolution Optimizer V 10.0.28) used for optimized algorithm of VAMT. A phantom for experiment virtually created at treatment planning system, 30x30x30 cm sized, homogeneous density (HU: 0) and heterogeneous density that inserted air assumed material (HU: -1,000). Apply to clinical treatment planning on the basis of general treatment planning feature analyzed with Phantom planning. In homogeneous density phantom, PBC and AAA show 65.2% PDD (6 MV, 10 cm) both, In heterogeneous density phantom, also show similar PDD value before meet with low density material, but they show different dose curve in air territory, PDD 10 cm showed 75%, 73% each after penetrate phantom. 3D treatment plan in same MU, AAA treatment planning shows low dose at Lung included area. 2D POP treatment plan with 15 MV of cervical vertebral region include trachea and lung area, Conformity Index (ICRU 62) is 0.95 in PBC calculation and 0.93 in AAA. DVO DVH and Dose calculation DVH are showed equal value in IMRT treatment plan. But AAA calculation shows lack of dose compared with DVO result which is satisfactory condition. Optimizing VMAT treatment plans using PRO II obtained results were satisfactory, but lower density area showed lack of dose in dose calculations. PRO III, but optimizing the dose calculation results were similar with optimized the same conditions once more. In this study, do not judge the rightness of the dose

Analysis of Radiation Treatment Planning by Dose Calculation and Optimization Algorithm

International Nuclear Information System (INIS)

Kim, Dae Sup; Yoon, In Ha; Lee, Woo Seok; Baek, Geum Mun

2012-01-01

Analyze the Effectiveness of Radiation Treatment Planning by dose calculation and optimization algorithm, apply consideration of actual treatment planning, and then suggest the best way to treatment planning protocol. The treatment planning system use Eclipse 10.0. (Varian, USA). PBC (Pencil Beam Convolution) and AAA (Anisotropic Analytical Algorithm) Apply to Dose calculation, DVO (Dose Volume Optimizer 10.0.28) used for optimized algorithm of Intensity Modulated Radiation Therapy (IMRT), PRO II (Progressive Resolution Optimizer V 8.9.17) and PRO III (Progressive Resolution Optimizer V 10.0.28) used for optimized algorithm of VAMT. A phantom for experiment virtually created at treatment planning system, 30x30x30 cm sized, homogeneous density (HU: 0) and heterogeneous density that inserted air assumed material (HU: -1,000). Apply to clinical treatment planning on the basis of general treatment planning feature analyzed with Phantom planning. In homogeneous density phantom, PBC and AAA show 65.2% PDD (6 MV, 10 cm) both, In heterogeneous density phantom, also show similar PDD value before meet with low density material, but they show different dose curve in air territory, PDD 10 cm showed 75%, 73% each after penetrate phantom. 3D treatment plan in same MU, AAA treatment planning shows low dose at Lung included area. 2D POP treatment plan with 15 MV of cervical vertebral region include trachea and lung area, Conformity Index (ICRU 62) is 0.95 in PBC calculation and 0.93 in AAA. DVO DVH and Dose calculation DVH are showed equal value in IMRT treatment plan. But AAA calculation shows lack of dose compared with DVO result which is satisfactory condition. Optimizing VMAT treatment plans using PRO II obtained results were satisfactory, but lower density area showed lack of dose in dose calculations. PRO III, but optimizing the dose calculation results were similar with optimized the same conditions once more. In this study, do not judge the rightness of the dose

Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

International Nuclear Information System (INIS)

Butkovich, T.R.

1980-05-01

A generic test of the geologic storage of spent-fuel assemblies is being made at Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canisters, preliminary calculations were made using a pair of existing finite-element codes, ADINA and ADINAT

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

Energy Technology Data Exchange (ETDEWEB)

Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)

2014-05-15

The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong

2014-01-01

The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

Coupled thermo-hydro-mechanical calculations of the water saturation phase of a KBS-3 deposition hole. Influence of hydraulic rock properties on the water saturation phase

International Nuclear Information System (INIS)

Boergesson, Lennart; Hernelind, J.

1999-12-01

The wetting process in deposition holes designed according to the KBS-3-concept has been simulated with finite element calculations of the thermo-hydro-mechanical processes in the buffer, backfill and surrounding rock. The buffer material has been modelled according to the preliminary material models developed for swelling clay. The properties of the rock have been varied in order to investigate the influence of the rock properties and the hydraulic conditions on the wetting processes. In the modelling of the test holes the permeability of the rock matrix, the water supply from the backfill, the water pressure in the surrounding rock, the permeability of the disturbed zone around the deposition hole, the water retention properties of the rock, and the transmissivity of two fractures intersecting the deposition hole have been varied. The calculations indicate that the wetting takes about 5 years if the water pressure in the rock is high and if the permeability of the rock is so high that the properties of the bentonite determine the wetting rate. However, it may take considerably more than 30 years if the rock is very tight and the water pressure in the rock is low. The calculations also show that the influence of the rock structure is rather large except for the influence of the transmissivity T of the fractures, which turned out to be insignificant for the values used in the calculations

Preliminary assessments the shortcut to remediation (category III-surplus facility assessments)

International Nuclear Information System (INIS)

Byars, L.L.

1995-01-01

This report presents the details of the preliminary assessments for the shortcut of decontamination of surplus nuclear facilities. Topics discussed include: environment, health and safety concerns; economic considerations; reduction of transition time; preliminary characterization reports; preliminary project plan; health and safety plan; quality assurance plan; surveillance and maintenance plan; and waste management plan

Preliminary study on the energy coefficients of buildings; Vorstudie zur Erhebung von Energiekennzahlen von Wohnbauten

Energy Technology Data Exchange (ETDEWEB)

Dettli, R.; Bade, S. [Econcept AG, Zuerich (Switzerland); Baumgartner, A.; Bleisch, M. [Amstein und Walthert, Zuerich (Switzerland)

2007-11-15

This comprehensive report for the Swiss Federal Office of Energy (SFOE) presents the results of a preliminary study concerning the definition of a method for the cost-effective and reliable collection of data and the calculation of energy coefficients for residential buildings in Switzerland. On the basis of data already collected, typical coefficients for various types of building are proposed. Also, reasons for considerable differences between the data of various Swiss Cantons are investigated. Requirements and criteria for the judgement of the energy coefficients are discussed and the methods used by various Swiss cities and Cantons are reviewed. A comprehensive appendix completes the report.

The genetic variability of the Podolica cattle breed from the Gargano area. Preliminary results

Directory of Open Access Journals (Sweden)

Dario Cianci

2010-01-01

Full Text Available The Podolica cattle breed is autochthonous of Southern Italy and denoted by its particular rusticity. This study presents the preliminary results of the genetic characterization of the Podolica breed using DNA STR markers. A total of 20 microsatellite loci were analysed in 79 individuals reared in the Gargano area. Number of polymorphisms, allele fre- quencies, deviations from Hardy-Weinberg proportions, linkage disequilibrium between loci and genetic similarities between animals were calculated. The results showed a high deficiency of heterozygotes, the observed mean of het- erozygosis being 0.449, whereas the expected mean was 0.766. Many markers showed also deviations from the Hardy- Weinberg proportions and significant linkage disequilibrium between loci. However the genetic similarity within the pop- ulation was low (0.281 and the average number of alleles per locus was high (10, representing a high genetic vari- ability. In order to explain these results, a stratification of the breed in sub-populations with a high interior genetic homo- geneity but markedly differentiated one from each other could be hypothesized; this situation probably derived from non- random mating within each herd (consanguinity and from the lack of exchange of genetic material between the herds. A further study is needed on a wider sample and extending the analysis to FAO-ISAG microsatellite panel in order to con- firm this hypothesis. This could eventually provide the information necessary for the correct management of the repro- ductive schemes and for genomic traceability of meat production.

Comparison between ASHRAE and ISO thermal transmittance calculation methods

DEFF Research Database (Denmark)

Blanusa, Petar; Goss, William P.; Roth, Hartwig

2007-01-01

is proportional to the glazing/frame sightline distance that is also proportional to the total glazing spacer length. An example calculation of the overall heat transfer and thermal transmittance (U-value or U-factor) using the two methods for a thermally broken, aluminum framed slider window is presented....... The fenestration thermal transmittance calculations analyses presented in this paper show that small differences exist between the calculated thermal transmittance values produced by the ISO and ASHRAE methods. The results also show that the overall thermal transmittance difference between the two methodologies...... decreases as the total window area (glazing plus frame) increases. Thus, the resulting difference in thermal transmittance values for the two methods is negligible for larger windows. This paper also shows algebraically that the differences between the ISO and ASHRAE methods turn out to be due to the way...

Reactor kinetics calculated in the summation method and key delayed-neutron data

International Nuclear Information System (INIS)

Oyamatsu, Kazuhiro

2001-01-01

The point-reactor kinetics after a step reactivity insertion to a critical condition is solved directly form fission-product (FP) data (fission yields and decay data) for the first time. Numerical calculations are performed with the FP data in ENDF/B-VI. The inhour equation obtained directly from the FP data shows a different behavior at long periods from the one obtained from Tuttle's six-group parameter sets. The behavior is quite similar to the one obtained from the six-group parameter sets in ENDF/B-VI, that were obtained from FP data in a preliminary version of ENDF/B-VI. To identify the erroneous FP data, we examine the asymptotic form of the inhour equation at an infinitely long period. It is found that the most important precursors for long reactor periods are found 137 I, 88 Br and 87 Br. They cover more than 60% of the reactivity. It is remarkable that 137 I alone covers 30-50% depending on the fissioning system. In addition to the three precursors, 136 Te is found a candidate precursor for the peculiarity from the time dependence of the delayed neutron activity. It is recommended that the precision of their Pn values should be improved experimentally. For 137 I, 88 Br, and 87 Br, the relative uncertainty, dPn/Pn, should be decreased down to 2% and for 136 Te to 5%. (author)

Achieving High Accuracy in Calculations of NMR Parameters

DEFF Research Database (Denmark)

Faber, Rasmus

quantum chemical methods have been developed, the calculation of NMR parameters with quantitative accuracy is far from trivial. In this thesis I address some of the issues that makes accurate calculation of NMR parameters so challenging, with the main focus on SSCCs. High accuracy quantum chemical......, but no programs were available to perform such calculations. As part of this thesis the CFOUR program has therefore been extended to allow the calculation of SSCCs using the CC3 method. CC3 calculations of SSCCs have then been performed for several molecules, including some difficult cases. These results show...... vibrations must be included. The calculation of vibrational corrections to NMR parameters has been reviewed as part of this thesis. A study of the basis set convergence of vibrational corrections to nuclear shielding constants has also been performed. The basis set error in vibrational correction...

Parameters calculation of fuel assembly with complex geometry

International Nuclear Information System (INIS)

Wu Hongchun; Ju Haitao; Yao Dong

2006-01-01

The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)

Improved SVR Model for Multi-Layer Buildup Factor Calculation

International Nuclear Information System (INIS)

Trontl, K.; Pevec, D.; Smuc, T.

2006-01-01

The accuracy of point kernel method applied in gamma ray dose rate calculations in shielding design and radiation safety analysis is limited by the accuracy of buildup factors used in calculations. Although buildup factors for single-layer shields are well defined and understood, buildup factors for stratified shields represent a complex physical problem that is hard to express in mathematical terms. The traditional approach for expressing buildup factors of multi-layer shields is through semi-empirical formulas obtained by fitting the results of transport theory or Monte Carlo calculations. Such an approach requires an ad-hoc definition of the fitting function and often results with numerous and usually inadequately explained and defined correction factors added to the final empirical formula. Even more, finally obtained formulas are generally limited to a small number of predefined combinations of materials within relatively small range of gamma ray energies and shield thicknesses. Recently, a new approach has been suggested by the authors involving one of machine learning techniques called Support Vector Machines, i.e., Support Vector Regression (SVR). Preliminary investigations performed for double-layer shields revealed great potential of the method, but also pointed out some drawbacks of the developed model, mostly related to the selection of one of the parameters describing the problem (material atomic number), and the method in which the model was designed to evolve during the learning process. It is the aim of this paper to introduce a new parameter (single material buildup factor) that is to replace the existing material atomic number as an input parameter. The comparison of two models generated by different input parameters has been performed. The second goal is to improve the evolution process of learning, i.e., the experimental computational procedure that provides a framework for automated construction of complex regression models of predefined

Partition calculation for zero-order and conjugate image removal in digital in-line holography.

Science.gov (United States)

Ma, Lihong; Wang, Hui; Li, Yong; Jin, Hongzhen

2012-01-16

Conventional digital in-line holography requires at least two phase-shifting holograms to reconstruct an original object without zero-order and conjugate image noise. We present a novel approach in which only one in-line hologram and two intensity values (namely the object wave intensity and the reference wave intensity) are required. First, by subtracting the two intensity values the zero-order diffraction can be completely eliminated. Then, an algorithm, called partition calculation, is proposed to numerically remove the conjugate image. A preliminary experimental result is given to confirm the proposed method. The method can simplify the procedure of phase-shifting digital holography and improve the practical feasibility for digital in-line holography.

Preliminary results of local earthquake tomography around Bali, Lombok, and Sumbawa regions

Energy Technology Data Exchange (ETDEWEB)

Nugraha, Andri Dian, E-mail: nugraha@gf.itb.ac.id; Puspito, Nanang T; Yudistira, Tedi [Global Geophysical Reserach Group, Faculty of Mining and Petroleum Engineering, Institute of Technology Bandung, JlGanesa 10, Bandung, 40132 (Indonesia); Kusnandar, Ridwan; Sakti, Artadi Pria [Meteorological, Climatological, and Geophysical Agency (MCGA) of Indonesian, Jakarta (Indonesia)

2015-04-24

Bali, Sumbawa, and Lombok regions are located in active tectonic influence by Indo-Australia plate subducts beneath Sunda plate in southern part and local back-arc thrust in northern part the region. Some active volcanoes also lie from eastern part of Java, Bali, Lombok and Sumbawa regions. Previous studies have conducted subsurface seismic velocity imaging using regional and global earthquake data around the region. In this study, we used P-arrival time from local earthquake networks compiled by MCGA, Indonesia within time periods of 2009 up to 2013 to determine seismic velocity structure and simultaneously hypocenter adjustment by applying seismic tomography inversion method. For the tomographic inversion procedure, we started from 1-D initial velocity structure. We evaluated the resolution of tomography inversion results through checkerboard test and calculating derivative weigh sum. The preliminary results of tomography inversion show fairly clearly high seismic velocity subducting Indo-Australian and low velocity anomaly around volcano regions. The relocated hypocenters seem to cluster around the local fault system such as back-arc thrust fault in northern part of the region and around local fault in Sumbawa regions. Our local earthquake tomography results demonstrated consistent with previous studies and improved the resolution. For future works, we will determine S-wave velocity structure using S-wave arrival time to enhance our understanding of geological processes and for much better interpretation.

The ArTéMiS wide-field sub-millimeter camera: preliminary on-sky performance at 350 microns

Science.gov (United States)

Revéret, Vincent; André, Philippe; Le Pennec, Jean; Talvard, Michel; Agnèse, Patrick; Arnaud, Agnès.; Clerc, Laurent; de Breuck, Carlos; Cigna, Jean-Charles; Delisle, Cyrille; Doumayrou, Eric; Duband, Lionel; Dubreuil, Didier; Dumaye, Luc; Ercolani, Eric; Gallais, Pascal; Groult, Elodie; Jourdan, Thierry; Leriche, Bernadette; Maffei, Bruno; Lortholary, Michel; Martignac, Jérôme; Rabaud, Wilfried; Relland, Johan; Rodriguez, Louis; Vandeneynde, Aurélie; Visticot, François

2014-07-01

ArTeMiS is a wide-field submillimeter camera operating at three wavelengths simultaneously (200, 350 and 450 μm). A preliminary version of the instrument equipped with the 350 μm focal plane, has been successfully installed and tested on APEX telescope in Chile during the 2013 and 2014 austral winters. This instrument is developed by CEA (Saclay and Grenoble, France), IAS (France) and University of Manchester (UK) in collaboration with ESO. We introduce the mechanical and optical design, as well as the cryogenics and electronics of the ArTéMiS camera. ArTeMiS detectors consist in Si:P:B bolometers arranged in 16×18 sub-arrays operating at 300 mK. These detectors are similar to the ones developed for the Herschel PACS photometer but they are adapted to the high optical load encountered at APEX site. Ultimately, ArTeMiS will contain 4 sub-arrays at 200 μm and 2×8 sub-arrays at 350 and 450 μm. We show preliminary lab measurements like the responsivity of the instrument to hot and cold loads illumination and NEP calculation. Details on the on-sky commissioning runs made in 2013 and 2014 at APEX are shown. We used planets (Mars, Saturn, Uranus) to determine the flat-field and to get the flux calibration. A pointing model was established in the first days of the runs. The average relative pointing accuracy is 3 arcsec. The beam at 350 μm has been estimated to be 8.5 arcsec, which is in good agreement with the beam of the 12 m APEX dish. Several observing modes have been tested, like "On- The-Fly" for beam-maps or large maps, spirals or raster of spirals for compact sources. With this preliminary version of ArTeMiS, we concluded that the mapping speed is already more than 5 times better than the previous 350 μm instrument at APEX. The median NEFD at 350 μm is 600 mJy.s1/2, with best values at 300 mJy.s1/2. The complete instrument with 5760 pixels and optimized settings will be installed during the first half of 2015.

Presentation of preliminary studies relative to the long duration disposal of medium level and long lived (MLLL) wastes

International Nuclear Information System (INIS)

Leroy, C.; Moreau, A.; Fayette, L.; Bellon, M.; Templier, J.C.; Macias, R.M.; Porcher, J.B.; Rey, F.; Hollender, F.; Girard, J.P.

2002-01-01

In the contract of objectives signed in 2001 with the government, the French atomic energy commission (CEA) committed itself to supply reports of preliminary studies about long duration disposal concepts for medium level and long lived radioactive wastes. This document makes the synthesis of the preliminary studies carried out in 2001 and 2002 by exploring simultaneously the surface and subsurface disposal concepts. The studies deal with the design of a facility with a long service life. Four hypotheses have been retained for the preliminary studies: a secular lifetime (typically 100 to 300 years), a single and new site for all waste packages (no existing facility available), two confinement barriers, an envelope-type site with specific characteristics (seismicity, climate conditions, airplane crash..). These preliminary studies show the existence of solutions for each option: with and without storage containers in both type (surface and subsurface) of facilities. They outline the necessity of studying more thoroughly some technical points. This instruction will be performed for the concepts retained after a multi-criteria analysis. (J.S.)

Preliminary study of a flux converter for experimental reactor

International Nuclear Information System (INIS)

Malouch, M.F.

1998-01-01

The purpose of this project is to define the characteristics of a flux converter dedicated to increase the fast neutron flux in irradiation devices placed in the core of Osiris experimental reactor. This preliminary work has dealt with the neutronic and thermal-hydraulic aspects of this problem. The synthesis of the results produced by the codes APOLLO2, DAIXY, MERCURE5.3 and FLICA-3M shows that a cylindrical converter equipped with 5 fissile rings can enhance the fast flux by a 35% factor in an experimental device set in its center. (A.C.)

Preliminary safety evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)

International Nuclear Information System (INIS)

Dunn, T.D.; Lommers, L.J.; Tangirala, V.E.

1994-04-01

A qualitative comparison between the safety characteristics of the Gas Turbine-Modular Helium Reactor (GT-MHR) and those of the steam cycle shows that the two designs achieve equivalent levels of overall safety performance. This comparison is obtained by applying the scaling laws to detailed steam-cycle computations as well as the conclusions obtained from preliminary GT-MHR model simulations. The gas turbine design is predicted to be superior for some event categories, while the steam cycle design is better for others. From a safety perspective, the GT-MHR has a modest advantage for pressurized conduction cooldown events. Recent computational simulations of 102 column, 550 MW(t) GT-MHR during a depressurized conduction cooldown show that peak fuel temperatures are within the limits. The GT-MHR has a significantly lower risk due to water ingress events under operating conditions. Two additional scenarios, namely loss of load event and turbine deblading event that are specific to the GT-MHR design are discussed. Preliminary evaluation of the GT-MHR's safety characteristics indicate that the GT-MHR can be expected to satisfy or exceed its safety requirements

Radiocesium and polonium in seals from the Baltic Sea. A preliminary study

International Nuclear Information System (INIS)

Holm, E.; Leisvik, M.

2002-01-01

137 Cs and 210 Po in muscle, liver and kidney from 5 seals from the Baltic Sea have been studied. Assuming that the seals consume 5% of their own body weight each day and by using previously obtained data for the doses in fish, the biological half time has been estimated to 20 days for Cesium and 35 days for Polonium. The maximum yearly doses are calculated to 76 μGy from 137 CS to muscle and 3 500 μGy to kidneys from 210 Po. These data are very preliminary, the number of individuals being so small, and it is difficult to estimate how much food other than fish the seals consume. Furthermore only few data for radioactive concentrations in whole fish exist, which especially impacts the results for Polonium. (LN)

37 CFR 1.480 - Demand for international preliminary examination.

Science.gov (United States)

2010-07-01

... 37 Patents, Trademarks, and Copyrights 1 2010-07-01 2010-07-01 false Demand for international... Provisions International Preliminary Examination § 1.480 Demand for international preliminary examination. (a) On the filing of a proper Demand in an application for which the United States International...

Review of Preliminary Analysis Techniques for Tension Structures.

Science.gov (United States)

1984-02-01

however,a linear dinamic analysis can be conducted for purposes of preliminary design, relative to the static configuration. It is noted that the amount of...16 Chapter 3. PRELIMINARY DESIGN OF TENSION STRUCTURES . . .. .. .. .... 22 S.3.1 Cable Systems . . . . . . . . . . . . .. .. .. .... 23...3.1.1 Singly-Connected Segments. .. .... ... 24 3.1.2 Multiply-Connected Segments . . .. .. .. .. 27 3.1.3 Linearized Dynamics of Cable Systems . . . . 29

The fifth Atomic Energy Research dynamic benchmark calculation with HEXTRAN-SMABRE

International Nuclear Information System (INIS)

Haenaelaeinen, Anitta

1998-01-01

The fifth Atomic Energy Research dynamic benchmark is the first Atomic Energy Research benchmark for coupling of the thermohydraulic codes and three-dimensional reactor dynamic core models. In VTT HEXTRAN 2.7 is used for the core dynamics and SMABRE 4.6 as a thermohydraulic model for the primary and secondary loops. The plant model for SMABRE is based mainly on two input models. the Loviisa model and standard WWER-440/213 plant model. The primary circuit includes six separate loops, totally 505 nodes and 652 junctions. The reactor pressure vessel is divided into six parallel channels. In HEXTRAN calculation 176 symmetry is used in the core. In the sequence of main steam header break at the hot standby state, the liquid temperature is decreased symmetrically in the core inlet which leads to return to power. In the benchmark, no isolations of the steam generators are assumed and the maximum core power is about 38 % of the nominal power at four minutes after the break opening in the HEXTRAN-SMABRE calculation. Due to boric acid in the high pressure safety injection water, the power finally starts to decrease. The break flow is pure steam in the HEXTRAN-SMABRE calculation during the whole transient even in the swell levels in the steam generators are very high due to flashing. Because of sudden peaks in the preliminary results of the steam generator heat transfer, the SMABRE drift-flux model was modified. The new model is a simplified version of the EPRI correlation based on test data. The modified correlation behaves smoothly. In the calculations nuclear data is based on the ENDF/B-IV library and it has been evaluated with the CASMO-HEX code. The importance of the nuclear data was illustrated by repeating the benchmark calculation with using three different data sets. Optimal extensive data valid from hot to cold conditions were not available for all types of fuel enrichments needed in this benchmark.(Author)

Preliminary Tests of a New Low-Cost Photogrammetric System

Science.gov (United States)

Santise, M.; Thoeni, K.; Roncella, R.; Sloan, S. W.; Giacomini, A.

2017-11-01

This paper presents preliminary tests of a new low-cost photogrammetric system for 4D modelling of large scale areas for civil engineering applications. The system consists of five stand-alone units. Each of the units is composed of a Raspberry Pi 2 Model B (RPi2B) single board computer connected to a PiCamera Module V2 (8 MP) and is powered by a 10 W solar panel. The acquisition of the images is performed automatically using Python scripts and the OpenCV library. Images are recorded at different times during the day and automatically uploaded onto a FTP server from where they can be accessed for processing. Preliminary tests and outcomes of the system are discussed in detail. The focus is on the performance assessment of the low-cost sensor and the quality evaluation of the digital surface models generated by the low-cost photogrammetric systems in the field under real test conditions. Two different test cases were set up in order to calibrate the low-cost photogrammetric system and to assess its performance. First comparisons with a TLS model show a good agreement.

WISE/NEOWISE OBSERVATIONS OF THE JOVIAN TROJANS: PRELIMINARY RESULTS

International Nuclear Information System (INIS)

Grav, T.; Mainzer, A. K.; Bauer, J.; Masiero, J.; Eisenhardt, P. R. M.; Blauvelt, E.; DeBaun, E.; Elsbury, D.; Gautier, T. IV; Gomillion, S.; Hand, E.; Wilkins, A.; Spahr, T.; McMillan, R. S.; Walker, R.; Cutri, R.; Wright, E.

2011-01-01

We present the preliminary analysis of over 1739 known and 349 candidate Jovian Trojans observed by the NEOWISE component of the Wide-field Infrared Survey Explorer (WISE). With this survey the available diameters, albedos, and beaming parameters for the Jovian Trojans have been increased by more than an order of magnitude compared to previous surveys. We find that the Jovian Trojan population is very homogenous for sizes larger than ∼10 km (close to the detection limit of WISE for these objects). The observed sample consists almost exclusively of low albedo objects, having a mean albedo value of 0.07 ± 0.03. The beaming parameter was also derived for a large fraction of the observed sample, and it is also very homogenous with an observed mean value of 0.88 ± 0.13. Preliminary debiasing of the survey shows that our observed sample is consistent with the leading cloud containing more objects than the trailing cloud. We estimate the fraction to be N(leading)/N(trailing) ∼ 1.4 ± 0.2, lower than the 1.6 ± 0.1 value derived by Szabó et al.

PRELIMINARY TESTS OF A NEW LOW-COST PHOTOGRAMMETRIC SYSTEM

Directory of Open Access Journals (Sweden)

M. Santise

2017-11-01

Full Text Available This paper presents preliminary tests of a new low-cost photogrammetric system for 4D modelling of large scale areas for civil engineering applications. The system consists of five stand-alone units. Each of the units is composed of a Raspberry Pi 2 Model B (RPi2B single board computer connected to a PiCamera Module V2 (8 MP and is powered by a 10 W solar panel. The acquisition of the images is performed automatically using Python scripts and the OpenCV library. Images are recorded at different times during the day and automatically uploaded onto a FTP server from where they can be accessed for processing. Preliminary tests and outcomes of the system are discussed in detail. The focus is on the performance assessment of the low-cost sensor and the quality evaluation of the digital surface models generated by the low-cost photogrammetric systems in the field under real test conditions. Two different test cases were set up in order to calibrate the low-cost photogrammetric system and to assess its performance. First comparisons with a TLS model show a good agreement.

Validation of an online risk calculator for the prediction of anastomotic leak after colon cancer surgery and preliminary exploration of artificial intelligence-based analytics.

Science.gov (United States)

Sammour, T; Cohen, L; Karunatillake, A I; Lewis, M; Lawrence, M J; Hunter, A; Moore, J W; Thomas, M L

2017-11-01

Recently published data support the use of a web-based risk calculator ( www.anastomoticleak.com ) for the prediction of anastomotic leak after colectomy. The aim of this study was to externally validate this calculator on a larger dataset. Consecutive adult patients undergoing elective or emergency colectomy for colon cancer at a single institution over a 9-year period were identified using the Binational Colorectal Cancer Audit database. Patients with a rectosigmoid cancer, an R2 resection, or a diverting ostomy were excluded. The primary outcome was anastomotic leak within 90 days as defined by previously published criteria. Area under receiver operating characteristic curve (AUROC) was derived and compared with that of the American College of Surgeons National Surgical Quality Improvement Program ® (ACS NSQIP) calculator and the colon leakage score (CLS) calculator for left colectomy. Commercially available artificial intelligence-based analytics software was used to further interrogate the prediction algorithm. A total of 626 patients were identified. Four hundred and fifty-six patients met the inclusion criteria, and 402 had complete data available for all the calculator variables (126 had a left colectomy). Laparoscopic surgery was performed in 39.6% and emergency surgery in 14.7%. The anastomotic leak rate was 7.2%, with 31.0% requiring reoperation. The anastomoticleak.com calculator was significantly predictive of leak and performed better than the ACS NSQIP calculator (AUROC 0.73 vs 0.58) and the CLS calculator (AUROC 0.96 vs 0.80) for left colectomy. Artificial intelligence-predictive analysis supported these findings and identified an improved prediction model. The anastomotic leak risk calculator is significantly predictive of anastomotic leak after colon cancer resection. Wider investigation of artificial intelligence-based analytics for risk prediction is warranted.

Many body calculations in atomic physics

International Nuclear Information System (INIS)

Kelly, H.P.

1985-01-01

The use of the many-body perturbation theory for atomic calculations are reviewed. The major emphasis is on the use of the linked-cluster many-body perturbation theory derived by Brueckner and Goldstone. Applications of many-body theory to calculations of hyperfine structure are examined. Auger rates and parity violation are discussed. Photoionization is reviewed, and the authors show how many-body perturbation theory can be applied to problems ranging from structural properties such as correlation energies and hyperfine structure to dynamical properties such as transitions induced by weak neutral currents and photoionization cross sections

Intracerebral metastasis showing restricted diffusion: Correlation with histopathologic findings

Energy Technology Data Exchange (ETDEWEB)

Duygulu, G. [Radiology Department, Ege University Medicine School, Izmir (Turkey); Ovali, G. Yilmaz [Radiology Department, Celal Bayar University Medicine School, Manisa (Turkey)], E-mail: gulgun.yilmaz@bayar.edu.tr; Calli, C.; Kitis, O.; Yuenten, N. [Radiology Department, Ege University Medicine School, Izmir (Turkey); Akalin, T. [Pathology Department, Ege University Medicine School, Izmir (Turkey); Islekel, S. [Neurosurgery Department, Ege University Medicine School, Izmir (Turkey)

2010-04-15

Objective: We aimed to detect the frequency of restricted diffusion in intracerebral metastases and to find whether there is correlation between the primary tumor pathology and diffusion-weighted MR imaging (DWI) findings of these metastases. Material and methods: 87 patients with intracerebral metastases were examined with routine MR imaging and DWI. 11 hemorrhagic metastatic lesions were excluded. The routine MR imaging included three plans before and after contrast enhancement. The DWI was performed with spin-echo EPI sequence with three b values (0, 500 and 1000), and ADC maps were calculated. 76 patients with metastases were grouped according to primary tumor histology and the ratios of restricted diffusion were calculated according to these groups. ADCmin values were measured within the solid components of the tumors and the ratio of metastases with restricted diffusion to that which do not show restricted diffusion were calculated. Fisher's exact and Mann-Whitney U tests were used for the statistical analysis. Results: Restricted diffusion was observed in a total of 15 metastatic lesions (19, 7%). Primary malignancy was lung carcinoma in 10 of these cases (66, 6%) (5 small cell carcinoma, 5 non-small cell carcinoma), and breast carcinoma in three cases (20%). Colon carcinoma and testicular teratocarcinoma were the other two primary tumors in which restricted diffusion in metastasis was detected. There was no statistical significant difference between the primary pathology groups which showed restricted diffusion (p > 0.05). ADCmin values of solid components of the metastasis with restricted diffusion and other metastasis without restricted diffusion also showed no significant statistical difference (0.72 {+-} 0.16 x 10{sup -3} mm{sup 2}/s and 0.78 {+-} 21 x 10{sup -3} mm{sup 2}/s respectively) (p = 0.325). Conclusion: Detection of restricted diffusion on DWI in intracerebral metastasis is not rare, particularly if the primary tumor is lung or breast

Preliminary report on the geology and metallogenesis of uranium in Bolivia

International Nuclear Information System (INIS)

Pardo-Leyton, E.; Barron, E.

1984-01-01

The paper is a preliminary study of aspects of the geology and metallogenesis of uranium in Bolivia. It must be considered as a first study which will serve as a basis for possible future work of this type. Its purpose is to show the possible relationships of the uranium anomalies and prospects to the country's various mineralized belts and also to the Nazca tectonic plate and to lineaments and intrusive bodies. (author)

Measurement of single kidney contrast media clearance by multiphasic spiral computed tomography: preliminary results

International Nuclear Information System (INIS)

Hackstein, N.; Puille, M.F.; Bak, Benjamin H.; Scharwat, Oliver; Rau, W.S.

2001-01-01

Objective. We present preliminary results of a new method (hereinafter called 'CT-clearance') to measure single kidney contrast media clearance by performing multiphasic helical CT of the kidneys. CT-clearance was calculated according to an extension of the Patlak-Plot. In contrast to prior investigators, who repeatedly measured a single slice, this method makes it possible to calculate single kidney clearance from at least three spiral CTs, utilizing the whole kidney volume. Methods. Spiral CT of the kidneys was performed unenhanced and about 30 and 100 s after administration of about 120 ml iopromide. Sum-density of the whole kidneys and aortic density was calculated from this data. Using this data, renal clearance of contrast media was calculated by CT-clearance in 29 patients. As reference, Serum-clearance was calculated in 24 patients by application of a modified one-exponential slope model. Information on the relative kidney function was gained by renal scintigraphy with Tc99m-MAG-3 or Tc99m-DMSA in 29 patients. Results. Linear regression analysis revealed a correlation coefficient of CT-clearance with Serum-clearance of r=0.78 with Cl (CT) [ml/min]=22.2+1.03 * Cl (serum), n=24. Linear regression of the relative kidney function (rkf) of the right kidney calculated by CT-clearance compared to scintigraphy results provided a correlation coefficient r=0.89 with rkf(CT)[%]=18.6+0.58 * rkf(scintigraphy), n=29. Conclusion. The obtained results of contrast media clearance measured by CT-clearance are in the physiological range of the parameter. Future studies should be performed to improve the methodology with the aim of higher accuracy. More specifically, better determination of the aortic density curve might improve the accuracy

A Novel Hybrid Similarity Calculation Model

Directory of Open Access Journals (Sweden)

Xiaoping Fan

2017-01-01

Full Text Available This paper addresses the problems of similarity calculation in the traditional recommendation algorithms of nearest neighbor collaborative filtering, especially the failure in describing dynamic user preference. Proceeding from the perspective of solving the problem of user interest drift, a new hybrid similarity calculation model is proposed in this paper. This model consists of two parts, on the one hand the model uses the function fitting to describe users’ rating behaviors and their rating preferences, and on the other hand it employs the Random Forest algorithm to take user attribute features into account. Furthermore, the paper combines the two parts to build a new hybrid similarity calculation model for user recommendation. Experimental results show that, for data sets of different size, the model’s prediction precision is higher than the traditional recommendation algorithms.

Pickering safeguards: a preliminary analysis

International Nuclear Information System (INIS)

Todd, J.L.; Hodgkinson, J.G.

1977-05-01

A summary is presented of thoughts relative to a systems approach for implementing international safeguards. Included is a preliminary analysis of the Pickering Generating Station followed by a suggested safeguards system for the facility

A preliminary assessment of potential doses to man from radioactive waste dumped in the Arctic Sea

Energy Technology Data Exchange (ETDEWEB)

Nielsen, S P [Forskningscente Risoe, Roskilde (Denmark); Iosjpe, M; Strand, P [Norwegian Radiation Protection Authority, Oesteraas (Norway)

1995-09-01

This report describes a preliminary radiological assessment of collective doses to the world population from radioactive material dumped in the Barents and Kara Seas in the period 1961-1991. Information on the dumped waste and the rates of release of radionuclides have been available from Russian sources and from the International Atomic Energy Agency. A box model has been used to simulate the dispersion of radionuclides in the marine environment and to calculate the contamination of seafood and the subsequent radiation doses to man. Two release scenarios have been adopted. The worst-case release scenario which ignores the presence of barriers between spent nuclear fuel and seawater is estimated to give rise to about 10 mansieverts calculated to 1000 years from the time of release. A more realistic release scenario is estimated to cause about 3 mansieverts. In both cases exposure from the radionuclide {sup 137}Cs is found to dominate the doses. (au) 8 tabs., 56 ills., 19 refs.

A preliminary assessment of potential doses to man from radioactive waste dumped in the Arctic sea

International Nuclear Information System (INIS)

Nielsen, S.P.; Iosjpe, M.; Strand, P.

1995-11-01

This report describes a preliminary radiological assessment of collective doses to the world population from radioactive material dumped in the Kara and Barents Seas in the period 1961-1991. Information on the dumped waste and the rates of release of radionuclides have been available from Russian sources and from the International Atomic Energy Agency. A box model has been used to simulate the dispersion of radionuclides in the marine environment and to calculate the contamination of seafood and the subsequent radiation doses to man. Two release scenarios have been adopted. The worst-case release scenario, which ignores the presence of barriers between spent nuclear fuel and seawater, is estimated to give rise to about 10 mansievert calculated to 1000 years from the time of release. A more realistic release scenario is estimated to cause about 3 mansieverts. In both cases exposure from the radionuclide 137 Cs is found to dominate the doses. 19 refs., 56 figs., 8 tabs

Alternative conceptual models and codes for unsaturated flow in fractured tuff: Preliminary assessments for GWTT-95

International Nuclear Information System (INIS)

Ho, C.K.; Altman, S.J.; Arnold, B.W.

1995-09-01

Groundwater travel time (GWTT) calculations will play an important role in addressing site-suitability criteria for the potential high-level nuclear waste repository at Yucca Mountain,Nevada. In support of these calculations, Preliminary assessments of the candidate codes and models are presented in this report. A series of benchmark studies have been designed to address important aspects of modeling flow through fractured media representative of flow at Yucca Mountain. Three codes (DUAL, FEHMN, and TOUGH 2) are compared in these benchmark studies. DUAL is a single-phase, isothermal, two-dimensional flow simulator based on the dual mixed finite element method. FEHMN is a nonisothermal, multiphase, multidimensional simulator based primarily on the finite element method. TOUGH2 is anon isothermal, multiphase, multidimensional simulator based on the integral finite difference method. Alternative conceptual models of fracture flow consisting of the equivalent continuum model (ECM) and the dual permeability (DK) model are used in the different codes

A preliminary assessment of potential doses to man from radioactive waste dumped in the Arctic Sea

International Nuclear Information System (INIS)

Nielsen, S.P.; Iosjpe, M.; Strand, P.

1995-09-01

This report describes a preliminary radiological assessment of collective doses to the world population from radioactive material dumped in the Barents and Kara Seas in the period 1961-1991. Information on the dumped waste and the rates of release of radionuclides have been available from Russian sources and from the International Atomic Energy Agency. A box model has been used to simulate the dispersion of radionuclides in the marine environment and to calculate the contamination of seafood and the subsequent radiation doses to man. Two release scenarios have been adopted. The worst-case release scenario which ignores the presence of barriers between spent nuclear fuel and seawater is estimated to give rise to about 10 mansieverts calculated to 1000 years from the time of release. A more realistic release scenario is estimated to cause about 3 mansieverts. In both cases exposure from the radionuclide 137 Cs is found to dominate the doses. (au) 8 tabs., 56 ills., 19 refs

First-principles calculations of novel materials

Science.gov (United States)

Sun, Jifeng

relatively modest band masses for both electrons and holes suggesting applications. Optical properties show a infrared-red absorption when doped. This could potentially be useful for combining wavelength filtering and transparent conducting functions. Furthermore, our defect calculations show that Ba 2TeO is intrinsically p-type conducting under Ba-poor condition. However, the spontaneous formation of the donor defects may constrain the p-type transport properties and would need to be addressed to enable applications. Chapter 4 mainly devotes to the thermoelectric properties of the famous phase change material, Ge2Sb2Te5 (GST). GST has been used in data storage for more than a decade because of their fast phase switching between metastable crystalline (cubic) and amorphous phases. It also exhibits interesting thermoelectric properties, and we did a systematic study on the two crystalline phases (hexagonal and cubic) and the amorphous phase. We found a high Seebeck coefficient with a broad doping concentrations for both n-type and p-type, at and below room temperatures (300 K) for both the cubic and amorphous phases. This finding will be of crucial interests in further understand the thermoelectric properties experimentally and find device applications in the ultimate goal. Several magnetic materials that involve lanthanide elements are reported in Chapter 5. First of all, the electronic and magnetic properties of the BaLn2O4 (Ln = La-Lu, Y) family compound are studied. The series has been synthesized for the first time in single crystalline form, using a molten metal flux. They crystallize in the CaV 2O4 structure type with primitive orthorhombic symmetry (space group Pnma, #62). Our calculations show an insulating character with band gaps ranging from 3 eV to 4.5 eV for the three representative compounds, BaLa2O4, BaGd2O4 and BaLu 2O4. Moreover, the superexchange magnetism is also studied. Secondly, a strong correlated system with cerium is investigated. As expected, we

Preliminary bioelectrical impedance analysis (BIA) equation for body composition assessment in young females from Colombia

International Nuclear Information System (INIS)

Caicedo, J C; González-Correa, C H; González-Correa, C A

2013-01-01

A previous study showed that reported BIA equations for body composition are not suitable for Colombian population. The purpose of this study was to develop and validate a preliminary BIA equation for body composition assessment in young females from Colombia, using hydrodensitometry as reference method. A sample of 30 young females was evaluated. Inclusion and exclusion criteria were defined to minimize the variability of BIA. Height, weight, BIA, residual lung volume (RV) and underwater weight (UWW) were measured. A preliminary BIA equation was developed (r 2 = 0.72, SEE = 2.48 kg) by stepwise multiple regression with fat-free mass (FFM) as dependent variable and weight, height and impedance measurements as independent variables. The quality of regression was evaluated and a cross-validation against 50% of sample confirmed that results obtained with the preliminary BIA equation is interchangeable with results obtained with hydrodensitometry (r 2 = 0.84, SEE = 2.62 kg). The preliminary BIA equation can be used for body composition assessment in young females from Colombia until a definitive equation is developed. The next step will be increasing the sample, including a second reference method, as deuterium oxide dilution (D 2 O), and using multi-frequency BIA (MF-BIA). It would also be desirable to develop equations for males and other ethnic groups in Colombia.

Preliminar calculation of tornado risk in the site of Ipero

Energy Technology Data Exchange (ETDEWEB)

Pacheco, Rafael R.; Costa, Saulo Barros, E-mail: rafael.rade@ctmsp.mar.mil.br, E-mail: saulo.costa@ctmsp.mar.mil.br [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Andrade, Delvonei A., E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

General Design Criterion (GDC) 2 to 10 CFR 50 requires that 'structures, systems, and components that are important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes, without loss of capability to perform their safety functions'. According to Regulatory Guide 1.76, the design-basis intensity of a tornado for a nuclear power plant shall not exceed the intensity of the strongest tornado that occurs with the frequency of 10-7/years. Reinforcing the plant to achieve this goal represents a high increase in the costs of the project, and correspondently increase in the time required to have it commissioned. This way, the right definition of tornado risk in a site would represent savings in money for the project and in time for the licensing of a nuclear power plants. This works aims to establish a preliminary calculation of the tornado risk in the site of Ipero, where will work LABGENE from Brazilian Navy, and RMB from CNEN. (author)

Preliminary I&C Design for LORELEI

International Nuclear Information System (INIS)

Korotkin, S.; Kaufman, Y.; Guttmann, E. B.; Levy, S.; Amidan, D.; Gdalyho, B.; Cahana, T.; Ellenbogen, A.; Arad, M.; Weiss, Y.; Sasson, A.; Ferry, L.; Bourrelly, F.; Cohen, Y.

2014-01-01

This document summarizes the preliminary I&C design for LORELEI experiment The preliminary design deals with considerations regarding appropriate safety and service instrumentation. The determined closed loop control rules for temperature and position will be implemented in the detailed design. The Computer Aided Operator Decisions System (CAODS) will be used for prediction of hot spot temperature and thickness of oxidation layer using Baker-Just correlation. The proposed hybrid simulation system comprising of both virtual and real hardware will be in-cooperated for LORELEI verification. It will perform both integration cold tests for a partial hardware loop and virtual tests for the final I&C design

Cell emulation and preliminary results.

Science.gov (United States)

2016-07-01

This report details preliminary results of the testing plan implemented by the Hawaii Natural Energy Institute to evaluate Electric Vehicle (EV) battery durability and reliability under electric utility grid operations. Commercial EV battery cells ar...

The preliminary design and feasibility study of the spent fuel and high level waste repository in the Czech Republic

International Nuclear Information System (INIS)

Valvoda, Z.; Holub, J.; Kucerka, M.

1996-01-01

In the year 1993, began the Program of Development of the Spent Fuel and High Level Waste Repository in the Conditions of the Czech Republic. During the first phase, the basic concept and structure of the Program has been developed, and the basic design criteria and requirements were prepared. In the conditions of the Czech Republic, only an underground repository in deep geological formation is acceptable. Expected depth is between 500 to 1000 meters and as host rock will be granites. A preliminary variant design study was realized in 1994, that analyzed the radioactive waste and spent fuel flow from NPPs to the repository, various possibilities of transportation in accordance to the various concepts of spent fuel conditioning and transportation to the underground structures. Conditioning and encapsulation of spent fuel and/or radioactive waste is proposed on the repository site. Underground disposal structures are proposed at one underground floor. The repository will have reserve capacity for radioactive waste from NPPs decommissioning and for waste non acceptable to other repositories. Vertical disposal of unshielded canisters in boreholes and/or horizontal disposal of shielded canisters is studied. As the base term of the start up of the repository operation, the year 2035 has been established. From this date, a preliminary time schedule of the Project has been developed. A method of calculating leveled and discounted costs within the repository lifetime, for each of selected 5 variants, was used for economic calculations. Preliminary expected parametric costs of the repository are about 0,1 Kc ($0.004) per MWh, produced in the Czech NPPs. In 1995, the design and feasibility study has gone in more details to the technical concept of repository construction and proposed technologies, as well as to the operational phase of the repository. Paper will describe results of the 1995 design work and will present the program of the repository development in next period

Relativistic mean field calculations in neutron-rich nuclei

Energy Technology Data Exchange (ETDEWEB)

Gangopadhyay, G.; Bhattacharya, Madhubrata [Department of Physics, University of Calcutta, 92 Acharya Prafulla Chandra Road, Kolkata 700 009 (India); Roy, Subinit [Saha Institute of Nuclear Physics, Block AF, Sector 1, Kolkata- 700 064 (India)

2014-08-14

Relativistic mean field calculations have been employed to study neutron rich nuclei. The Lagrange's equations have been solved in the co-ordinate space. The effect of the continuum has been effectively taken into account through the method of resonant continuum. It is found that BCS approximation performs as well as a more involved Relativistic Continuum Hartree Bogoliubov approach. Calculations reveal the possibility of modification of magic numbers in neutron rich nuclei. Calculation for low energy proton scattering cross sections shows that the present approach reproduces the density in very light neutron rich nuclei.

Propagation calculation for reactor cases

Energy Technology Data Exchange (ETDEWEB)

Yang Yanhua [School of Power and Energy Engineering, Shanghai Jiao Tong Univ., Shanghai (China); Moriyama, K.; Maruyama, Y.; Nakamura, H.; Hashimoto, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-11-01

The propagation of steam explosion for real reactor geometry and conditions are investigated by using the computer code JASMINE-pro. The ex-vessel steam explosion is considered, which is described as follow: during the accident of reactor core meltdown, the molten core melts a hole at the bottom of reactor vessel and causes the higher temperature core fuel being leaked into the water pool below reactor vessel. During the melt-water mixing interaction process, the high temperature melt evaporates the cool water at an extreme high rate and might induce a steam explosion. A steam explosion could experience first the premixing phase and then the propagation explosion phase. For a propagation calculation, we should know the information about the initial fragmentation time, the total melt mass, premixing region size, initial void fraction and distribution of the melt volume fraction, and so on. All the initial conditions used in this calculation are based on analyses from some simple assumptions and the observation from the experiments. The results show that the most important parameter for the initial condition of this phase is the total mass and its initial distribution. This gives the requirement for a premixing calculation. On the other hand, for higher melt volume fraction case, the fragmentation is strong so that the local pressure can exceed over the EOS maximum pressure of the code, which lead to the incorrect calculation or divergence of the calculation. (Suetake, M.)

Digital assessment of preliminary impression accuracy for edentulous jaws: Comparisons of 3-dimensional surfaces between study and working casts.

Science.gov (United States)

Matsuda, Takashi; Goto, Takaharu; Kurahashi, Kosuke; Kashiwabara, Toshiya; Watanabe, Megumi; Tomotake, Yoritoki; Nagao, Kan; Ichikawa, Tetsuo

2016-07-01

The aim of this study was to compare 3-dimensional surfaces of study and working casts for edentulous jaws and to evaluate the accuracy of preliminary impressions with a view to the future application of digital dentistry for edentulous jaws. Forty edentulous volunteers were serially recruited. Nine dentists took preliminary and final impressions in a routine clinical work-up. The study and working casts were digitized using a dental 3-dimensional scanner. The two surface images were superimposed through a least-square algorithm using imaging software and compared qualitatively. Furthermore, the surface of each jaw was divided into 6 sections, and the difference between the 2 images was quantitatively evaluated. Overall inspection showed that the difference around residual ridges was small and that around borders were large. The mean differences in the upper and lower jaws were 0.26mm and 0.45mm, respectively. The maximum values of the differences showed that the upward change mainly occurred in the anterior residual ridge, and the downward change mainly in the posterior border seal, and the labial and buccal vestibules, whereas every border of final impression was shortened in the lower jaw. The accuracy in all areas except the border, which forms the foundation, was estimated to be less than 0.25mm. Using digital technology, we here showed the overall and sectional accuracy of the preliminary impression for edentulous jaws. In our clinic, preliminary impressions have been made using an alginate material while ensuring that the requisite impression area was covered. Copyright © 2016 Japan Prosthodontic Society. Published by Elsevier Ltd. All rights reserved.

Study on the concurrent calculation with a multitransputer network

International Nuclear Information System (INIS)

Inomata, Shinji; Suzuki, Katsuo

1992-06-01

A relation between the calculation processing time and the overhead of multitransputer network is studied in performing several concurrent calculations. Hardware configuration we used was a ring-type linked nine transputer network on an INMOS B008 mother board which was inserted into a slot of a personal computer compatible with IBM PC/AT. Further we used the occam2 toolset of MS-DOS version to write the calculation programs in the occam2 language. On arithmetic operations, some trigonometric and transcendental functions and numerical integrations, the calculation time and overhead were measured. The results show that the relative calculation speed is proportional to the number of transputers when the ratio of calculation time to overhead is much greater than 1.0. And the overhead is consumed in inter-communicating among transputers for the parallel calculation set-up. (author)

Application of the Firefly and Luus-Jaakola algorithms in the calculation of a double reactive azeotrope

Science.gov (United States)

Mendes Platt, Gustavo; Pinheiro Domingos, Roberto; Oliveira de Andrade, Matheus

2014-01-01

The calculation of reactive azeotropes is an important task in the preliminary design and simulation of reactive distillation columns. Classically, homogeneous nonreactive azeotropes are vapor-liquid coexistence conditions where phase compositions are equal. For homogeneous reactive azeotropes, simultaneous phase and chemical equilibria occur concomitantly with equality of compositions (in the Ung-Doherty transformed space). The modeling of reactive azeotrope calculation is represented by a nonlinear algebraic system with phase equilibrium, chemical equilibrium and azeotropy equations. This nonlinear system can exhibit more than one solution, corresponding to a double reactive azeotrope. In a previous paper (Platt et al 2013 J. Phys.: Conf. Ser. 410 012020), we investigated some numerical aspects of the calculation of reactive azeotropes in the isobutene + methanol + methyl-tert-butyl-ether (with two reactive azeotropes) system using two metaheuristics: the Luus-Jaakola adaptive random search and the Firefly algorithm. Here, we use a hybrid structure (stochastic + deterministic) in order to produce accurate results for both azeotropes. After identifying the neighborhood of the reactive azeotrope, the nonlinear algebraic system is solved using Newton's method. The results indicate that using metaheuristics and some techniques devoted to the calculation of multiple minima allows both azeotropic coordinates in this reactive system to be obtains. In this sense, we provide a comprehensive analysis of a useful framework devoted to solving nonlinear systems, particularly in phase equilibrium problems.

Application of the Firefly and Luus–Jaakola algorithms in the calculation of a double reactive azeotrope

International Nuclear Information System (INIS)

Platt, Gustavo Mendes; Domingos, Roberto Pinheiro; Andrade, Matheus Oliveira de

2014-01-01

The calculation of reactive azeotropes is an important task in the preliminary design and simulation of reactive distillation columns. Classically, homogeneous nonreactive azeotropes are vapor–liquid coexistence conditions where phase compositions are equal. For homogeneous reactive azeotropes, simultaneous phase and chemical equilibria occur concomitantly with equality of compositions (in the Ung–Doherty transformed space). The modeling of reactive azeotrope calculation is represented by a nonlinear algebraic system with phase equilibrium, chemical equilibrium and azeotropy equations. This nonlinear system can exhibit more than one solution, corresponding to a double reactive azeotrope. In a previous paper (Platt et al 2013 J. Phys.: Conf. Ser. 410 012020), we investigated some numerical aspects of the calculation of reactive azeotropes in the isobutene + methanol + methyl-tert-butyl-ether (with two reactive azeotropes) system using two metaheuristics: the Luus–Jaakola adaptive random search and the Firefly algorithm. Here, we use a hybrid structure (stochastic + deterministic) in order to produce accurate results for both azeotropes. After identifying the neighborhood of the reactive azeotrope, the nonlinear algebraic system is solved using Newton's method. The results indicate that using metaheuristics and some techniques devoted to the calculation of multiple minima allows both azeotropic coordinates in this reactive system to be obtains. In this sense, we provide a comprehensive analysis of a useful framework devoted to solving nonlinear systems, particularly in phase equilibrium problems. (paper)

Benchmark calculations on resonance absorption by 238U in a PWR pin-cell geometry

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Janssen, A.J.

1993-12-01

Very accurate Monte Carlo calculations with MCNP have been performed to serve as a reference for benchmark calculations on resonance absorption by 238 U in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code ROLAIDS-CPM show that this code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (orig.)

Preliminary Results on Uncertainty Quantification for Pattern Analytics

Energy Technology Data Exchange (ETDEWEB)

Stracuzzi, David John [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Brost, Randolph [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Chen, Maximillian Gene [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Malinas, Rebecca [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Peterson, Matthew Gregor [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Phillips, Cynthia A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Robinson, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Woodbridge, Diane [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2015-09-01

This report summarizes preliminary research into uncertainty quantification for pattern ana- lytics within the context of the Pattern Analytics to Support High-Performance Exploitation and Reasoning (PANTHER) project. The primary focus of PANTHER was to make large quantities of remote sensing data searchable by analysts. The work described in this re- port adds nuance to both the initial data preparation steps and the search process. Search queries are transformed from does the specified pattern exist in the data? to how certain is the system that the returned results match the query? We show example results for both data processing and search, and discuss a number of possible improvements for each.

Application of MCNP in the criticality calculation for reactors

International Nuclear Information System (INIS)

Zhong Zhaopeng; Shi Gong; Hu Yongming

2003-01-01

The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

Preliminary Computational Analysis of the (HIRENASD) Configuration in Preparation for the Aeroelastic Prediction Workshop

Science.gov (United States)

Chwalowski, Pawel; Florance, Jennifer P.; Heeg, Jennifer; Wieseman, Carol D.; Perry, Boyd P.

2011-01-01

This paper presents preliminary computational aeroelastic analysis results generated in preparation for the first Aeroelastic Prediction Workshop (AePW). These results were produced using FUN3D software developed at NASA Langley and are compared against the experimental data generated during the HIgh REynolds Number Aero- Structural Dynamics (HIRENASD) Project. The HIRENASD wind-tunnel model was tested in the European Transonic Windtunnel in 2006 by Aachen University0s Department of Mechanics with funding from the German Research Foundation. The computational effort discussed here was performed (1) to obtain a preliminary assessment of the ability of the FUN3D code to accurately compute physical quantities experimentally measured on the HIRENASD model and (2) to translate the lessons learned from the FUN3D analysis of HIRENASD into a set of initial guidelines for the first AePW, which includes test cases for the HIRENASD model and its experimental data set. This paper compares the computational and experimental results obtained at Mach 0.8 for a Reynolds number of 7 million based on chord, corresponding to the HIRENASD test conditions No. 132 and No. 159. Aerodynamic loads and static aeroelastic displacements are compared at two levels of the grid resolution. Harmonic perturbation numerical results are compared with the experimental data using the magnitude and phase relationship between pressure coefficients and displacement. A dynamic aeroelastic numerical calculation is presented at one wind-tunnel condition in the form of the time history of the generalized displacements. Additional FUN3D validation results are also presented for the AGARD 445.6 wing data set. This wing was tested in the Transonic Dynamics Tunnel and is commonly used in the preliminary benchmarking of computational aeroelastic software.

Cloning, expression, purification, crystallization and preliminary X-ray crystallographic study of DHNA synthetase from Geobacillus kaustophilus

International Nuclear Information System (INIS)

Kanaujia, Shankar Prasad; Ranjani, Chellamuthu Vasuki; Jeyakanthan, Jeyaraman; Baba, Seiki; Kuroishi, Chizu; Ebihara, Akio; Shinkai, Akeo; Kuramitsu, Seiki; Shiro, Yoshitsugu; Sekar, Kanagaraj; Yokoyama, Shigeyuki

2007-01-01

DHNA synthetase from G. kaustophilus has been cloned, expressed, purified and crystallized. The aerobic Gram-positive bacterium Geobacillus kaustophilus is a bacillus species that was isolated from deep-sea sediment from the Mariana Trench. 1,4-Dihydroxy-2-naphthoate (DHNA) synthetase plays a vital role in the biosynthesis of menaquinone (vitamin K 2 ) in this bacterium. DHNA synthetase from Geobacillus kaustophilus was crystallized in the orthorhombic space group C222 1 , with unit-cell parameters a = 77.01, b = 130.66, c = 131.69 Å. The crystal diffracted to a resolution of 2.2 Å. Preliminary studies and molecular-replacement calculations reveal the presence of three monomers in the asymmetric unit

Preliminary Assessment of Detection Efficiency for the Geostationary Lightning Mapper Using Intercomparisons with Ground-Based Systems

Science.gov (United States)

Bateman, Monte; Mach, Douglas; Blakeslee, Richard J.; Koshak, William

2018-01-01

As part of the calibration/validation (cal/val) effort for the Geostationary Lightning Mapper (GLM) on GOES-16, we need to assess instrument performance (detection efficiency and accuracy). One major effort is to calculate the detection efficiency of GLM by comparing to multiple ground-based systems. These comparisons will be done pair-wise between GLM and each other source. A complication in this process is that the ground-based systems sense different properties of the lightning signal than does GLM (e.g., RF vs. optical). Also, each system has a different time and space resolution and accuracy. Preliminary results indicate that GLM is performing at or above its specification.

Preliminary observations of gate valve flow interruption tests, Phase 2

International Nuclear Information System (INIS)

Steele, R. Jr.; DeWall, K.G.

1990-01-01

This paper presents preliminary observations from the US Nuclear Regulatory Commission/Idaho National Engineering Laboratory Flexible Wedge Gate Valve Qualification and Flow Interruption Test Program, Phase 2. The program investigated the ability of selected boiling water reactor (BWR) process line valves to perform their containment isolation function at high energy pipe break conditions and other more normal flow conditions. The fluid and valve operating responses were measured to provide information concerning valve and operator performance at various valve loadings so that the information could be used to assess typical nuclear industry motor operator sizing equations. Six valves were tested, three 6-in. isolation valves representative of those used in reactor water cleanup systems in BWRs and three 10-in. isolation valves representative of those used in BWR high pressure coolant injection (HPCI) steam lines. The concern with these normally open isolation valves is whether they will close in the event of a downstream pipe break outside of containment. The results of this testing will provide part of the technical insights for NRC efforts regarding Generic Issue 87 (GI-87), Failure of the HPCI Steam Line Without Isolation, which includes concerns about the uncertainties in gate valve motor operator sizing and torque switch settings for these BWR containment isolation valves. As of this writing, the Phase 2 test program has just been completed. Preliminary observations made in the field confirmed most of the results from the Phase 1 test program. All six valves closing in high energy water, high energy steam, and high pressure cold water require more force to close than would be calculated using the typical variables in the standard industry motor operator sizing equations

Real-time POD-CFD Wind-Load Calculator for PV Systems

Energy Technology Data Exchange (ETDEWEB)

Huayamave, Victor [Centecorp; Divo, Eduardo [Centecorp; Ceballos, Andres [Centecorp; Barriento, Carolina [Centecorp; Stephen, Barkaszi [FSEC; Hubert, Seigneur [FSEC

2014-03-21

The primary objective of this project is to create an accurate web-based real-time wind-load calculator. This is of paramount importance for (1) the rapid and accurate assessments of the uplift and downforce loads on a PV mounting system, (2) identifying viable solutions from available mounting systems, and therefore helping reduce the cost of mounting hardware and installation. Wind loading calculations for structures are currently performed according to the American Society of Civil Engineers/ Structural Engineering Institute Standard ASCE/SEI 7; the values in this standard were calculated from simplified models that do not necessarily take into account relevant characteristics such as those from full 3D effects, end effects, turbulence generation and dissipation, as well as minor effects derived from shear forces on installation brackets and other accessories. This standard does not include provisions that address the special requirements of rooftop PV systems, and attempts to apply this standard may lead to significant design errors as wind loads are incorrectly estimated. Therefore, an accurate calculator would be of paramount importance for the preliminary assessments of the uplift and downforce loads on a PV mounting system, identifying viable solutions from available mounting systems, and therefore helping reduce the cost of the mounting system and installation. The challenge is that although a full-fledged three-dimensional computational fluid dynamics (CFD) analysis would properly and accurately capture the complete physical effects of air flow over PV systems, it would be impractical for this tool, which is intended to be a real-time web-based calculator. CFD routinely requires enormous computation times to arrive at solutions that can be deemed accurate and grid-independent even in powerful and massively parallel computer platforms. This work is expected not only to accelerate solar deployment nationwide, but also help reach the SunShot Initiative goals

V and V Efforts of Auroral Precipitation Models: Preliminary Results

Science.gov (United States)

Zheng, Yihua; Kuznetsova, Masha; Rastaetter, Lutz; Hesse, Michael

2011-01-01

Auroral precipitation models have been valuable both in terms of space weather applications and space science research. Yet very limited testing has been performed regarding model performance. A variety of auroral models are available, including empirical models that are parameterized by geomagnetic indices or upstream solar wind conditions, now casting models that are based on satellite observations, or those derived from physics-based, coupled global models. In this presentation, we will show our preliminary results regarding V&V efforts of some of the models.

A Preliminary Investigation into Parents' Concerns about Programming Education in Japanese Primary Schools

Science.gov (United States)

Maruyama, Yukiko; Kanoh, Hiroko; Adachi, Kinya

2017-01-01

To investigate parents' concerns about programming education in primary school, a preliminary online survey was carried out as a first step of the study. The result of the survey shows that parents seem to think that aim of programming education in primary school is not only learning coding. [For the complete proceedings, see ED579395.

Compilation report of VHTRC temperature coefficient benchmark calculations

Energy Technology Data Exchange (ETDEWEB)

Yasuda, Hideshi; Yamane, Tsuyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1995-11-01

A calculational benchmark problem has been proposed by JAERI to an IAEA Coordinated Research Program, `Verification of Safety Related Neutronic Calculation for Low-enriched Gas-cooled Reactors` to investigate the accuracy of calculation results obtained by using codes of the participating countries. This benchmark is made on the basis of assembly heating experiments at a pin-in block type critical assembly, VHTRC. Requested calculation items are the cell parameters, effective multiplication factor, temperature coefficient of reactivity, reaction rates, fission rate distribution, etc. Seven institutions from five countries have joined the benchmark works. Calculation results are summarized in this report with some remarks by the authors. Each institute analyzed the problem by applying the calculation code system which was prepared for the HTGR development of individual country. The values of the most important parameter, k{sub eff}, by all institutes showed good agreement with each other and with the experimental ones within 1%. The temperature coefficient agreed within 13%. The values of several cell parameters calculated by several institutes did not agree with the other`s ones. It will be necessary to check the calculation conditions again for getting better agreement. (J.P.N.).

Preliminary Uncertainty Analysis for SMART Digital Core Protection and Monitoring System

International Nuclear Information System (INIS)

Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun

2012-01-01

The Korea Atomic Energy Research Institute (KAERI) developed on-line digital core protection and monitoring systems, called SCOPS and SCOMS as a part of SMART plant protection and monitoring system. SCOPS simplified the protection system by directly connecting the four RSPT signals to each core protection channel and eliminated the control element assembly calculator (CEAC) hardware. SCOMS adopted DPCM3D method in synthesizing core power distribution instead of Fourier expansion method being used in conventional PWRs. The DPCM3D method produces a synthetic 3-D power distribution by coupling a neutronics code and measured in-core detector signals. The overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system was developed. In this paper, preliminary overall uncertainty factors for SCOPS/SCOMS of SMART initial core were evaluated by applying newly developed uncertainty analysis method

Adjoint spectrum calculation in fuel heterogeneous cells

International Nuclear Information System (INIS)

Suster, Luis Carlos

1998-01-01

In most codes for cells calculation, the multigroup cross sections are generated taking into consideration the conservation of the reaction rates in the forward spectrum. However, for certain uses of the perturbation theory it's necessary to use the average of the parameters for energy macrogroups over the forward and the adjoint spectra. In this thesis the adjoint spectrum was calculated from the adjoint neutron balance equations, that were obtained for a heterogeneous unit cell. The collision probabilities method was used to obtain these equations. In order optimize the computational run-time, the Gaussian quadrature method was used in the calculation of the neutron balance equations, forward and adjoint. This method of integration was also used for the Doppler broadening functions calculation, necessary for obtaining the energy dependent cross sections. In order to calculate the reaction rates and the average cross sections, using both the forward and the adjoint neutron spectra, the most important resonances of the U 238 were considered. The results obtained with the method show significant differences for the different cross sections weighting schemes. (author)

Researches on Preliminary Chemical Reactions in Spark-Ignition Engines

Science.gov (United States)

1943-06-01

compression type, without ignition, the resulting preliminary reactions being detectable and meas- urable thermometrically . Contents I. Influence of Preliminary...thoroughly insulated be- tween the carburettor and the engine, by aluminium foil and asbestos. -I -I " I" I ’I il i~ " !, I I 1𔃻I I’ ) To enable the

Head-and-neck IMRT treatments assessed with a Monte Carlo dose calculation engine

International Nuclear Information System (INIS)

Seco, J; Adams, E; Bidmead, M; Partridge, M; Verhaegen, F

2005-01-01

IMRT is frequently used in the head-and-neck region, which contains materials of widely differing densities (soft tissue, bone, air-cavities). Conventional methods of dose computation for these complex, inhomogeneous IMRT cases involve significant approximations. In the present work, a methodology for the development, commissioning and implementation of a Monte Carlo (MC) dose calculation engine for intensity modulated radiotherapy (MC-IMRT) is proposed which can be used by radiotherapy centres interested in developing MC-IMRT capabilities for research or clinical evaluations. The method proposes three levels for developing, commissioning and maintaining a MC-IMRT dose calculation engine: (a) development of a MC model of the linear accelerator, (b) validation of MC model for IMRT and (c) periodic quality assurance (QA) of the MC-IMRT system. The first step, level (a), in developing an MC-IMRT system is to build a model of the linac that correctly predicts standard open field measurements for percentage depth-dose and off-axis ratios. Validation of MC-IMRT, level (b), can be performed in a rando phantom and in a homogeneous water equivalent phantom. Ultimately, periodic quality assurance of the MC-IMRT system is needed to verify the MC-IMRT dose calculation system, level (c). Once the MC-IMRT dose calculation system is commissioned it can be applied to more complex clinical IMRT treatments. The MC-IMRT system implemented at the Royal Marsden Hospital was used for IMRT calculations for a patient undergoing treatment for primary disease with nodal involvement in the head-and-neck region (primary treated to 65 Gy and nodes to 54 Gy), while sparing the spinal cord, brain stem and parotid glands. Preliminary MC results predict a decrease of approximately 1-2 Gy in the median dose of both the primary tumour and nodal volumes (compared with both pencil beam and collapsed cone). This is possibly due to the large air-cavity (the larynx of the patient) situated in the centre

Pellet by pellet neutron flux calculations coupled with nodal expansion method

International Nuclear Information System (INIS)

Aldo, Dall'Osso

2003-01-01

We present a technique whose aim is to replace 2-dimensional pin by pin de-homogenization, currently done in core reactor calculations with the nodal expansion method (NEM), by a 3-dimensional finite difference diffusion calculation. This fine calculation is performed as a zoom in each node taking as boundary conditions the results of the NEM calculations. The size of fine mesh is of the order of a fuel pellet. The coupling between fine and NEM calculations is realised by an albedo like boundary condition. Some examples are presented showing fine neutron flux shape near control rods or assembly grids. Other fine flux behaviour as the thermal flux rise in the fuel near the reflector is emphasised. In general the results show the interest of the method in conditions where the separability of radial and axial directions is not granted. (author)

Preliminary result on trabecular bone score (TBS in lumbar vertebrae with experimentally altered microarchitecture

Directory of Open Access Journals (Sweden)

M. Di Stefano

2013-01-01

Full Text Available The aim of this preliminary research is to investigate the reliability of a new qualitative parameter, called Trabecular Bone Score (TBS, recently proposed for evaluating the microarchitectural arrangement of cancellous bone in scans carried out by dual energy X-ray absorptiometry (DXA. Vertebral bodies of 15 fresh samples of lumbar spines of adult pig were analysed either in basal conditions and with altered microarchitecture of the cancellous bone obtained by progressive drilling. The examined bony areas do not show changes in bone mineral density (BMD, whereas TBS values decrease with the increasing alteration of the vertebral microtrabecular structure. Our preliminary data seem to confirm the reliability of TBS as a qualitative parameter useful for evaluating the microarchitectural strength in bony areas quantitatively analysed by DXA.

A preliminary assessment of radiation effects on American Flagfish

Energy Technology Data Exchange (ETDEWEB)

Tzivaki, M.; Waller, E., E-mail: margarita.tzivaki@uoit.ca [University of Ontario Institute of Technology, Oshawa, ON (Canada)

2015-07-01

In order to add to the knowledge base of radiation effects on non-human biota, it is important to define benchmark values for different species. An experimental set-up was designed to investigate effects from irradiation with Cs-137 to American Flagfish. Preliminary experiments to assess the suitability of the methodology were conducted by exposing Flagfish eggs to 44 h of ionizing radiation. The subsequent observation of the developing fry showed no effect on hatching. However, the mortality and observed vertebral malformations were increased with increasing absorbed dose which is suspected to be a result of developmental defects in the embryonic stage. (author)

Preliminary study of the charged particle radiaton for th satellite power system

International Nuclear Information System (INIS)

Stassinopoulos, E.G.

1978-01-01

A preliminary radiation study was performed for the SPS project in order to determine the energetic charged particle environment for the three major phases of an SPS mission: the low earth orbit (LEO), the transfer ellipse (TE), and the synchronous geostationary trajectory (GEO). For that purpose, extensive calculations were performed and a large data base was generated, processeed, and analyzed. The external (surface incident) charged particle intensities, predicted for the SPS in each mission phase, were determined by orbital flux integration from the latest environment models. Magnetic field definitions for the three trajectories were obtained from a current field model. Spatial and temporal variations or conditions were considered and accounted for, where possible. Limited shielding and dose evaluations were performed for a simple geometry. The results of this analysis are presented in tabular and graphical form

Transplanting the Body: Preliminary Ethical Considerations.

Science.gov (United States)

Miller, Lantz Fleming

2017-11-01

A dissociated area of medical research warrants bioethical consideration: a proposed transplantation of a donor's entire body, except head, to a patient with a fatal degenerative disease. The seeming improbability of such an operation can only underscore the need for thorough bioethical assessment: Not assessing a case of such potential ethical import, by showing neglect instead of facing the issue, can only compound the ethical predicament, perhaps eroding public trust in ethical medicine. This article discusses the historical background of full-body transplantation, documents the seriousness of its current pursuit, and builds an argument for why prima facie this type of transplant is bioethically distinct. Certainly, this examination can only be preliminary, indicating what should be a wide and vigorous discussion among practitioners and ethicists. It concludes with practical suggestions for how the medical and bioethics community may proceed with ethical assessment.

Preliminary geothermal investigations at Manley Hot Springs, Alaska

Energy Technology Data Exchange (ETDEWEB)

East, J.

1982-04-01

Manley Hot Springs is one of several hot springs which form a belt extending from the Seward Peninsula to east-central Alaska. All of the hot springs are low-temperature, water-dominated geothermal systems, having formed as the result of circulation of meteoric water along deepseated fractures near or within granitic intrusives. Shallow, thermally disturbed ground at Manley Hot Springs constitutes an area of 1.2 km by 0.6 km along the lower slopes of Bean Ridge on the north side of the Tanana Valley. This area includes 32 springs and seeps and one warm (29.1/sup 0/C) well. The hottest springs range in temperature from 61/sup 0/ to 47/sup 0/C and are presently utilized for space heating and irrigation. This study was designed to characterize the geothermal system present at Manley Hot Springs and delineate likely sites for geothermal drilling. Several surveys were conducted over a grid system which included shallow ground temperature, helium soil gas, mercury soil and resistivity surveys. In addition, a reconnaissance ground temperature survey and water chemistry sampling program was undertaken. The preliminary results, including some preliminary water chemistry, show that shallow hydrothermal activity can be delineated by many of the surveys. Three localities are targeted as likely geothermal well sites, and a model is proposed for the geothermal system at Manley Hot Springs.

A Preliminary Review of Fatigue Among Rail Staff

OpenAIRE

Jialin Fan; Andrew P. Smith

2018-01-01

Background: Fatigue is a severe problem in the rail industry, which may jeopardize train crew's health and safety. Nonetheless, a preliminary review of all empirical evidence for train crew fatigue is still lacking. The aim of the present paper is, therefore, to provide a preliminary description of occupational fatigue in the rail industry. This paper reviews the literature with the research question examining the risk factors associated with train crew fatigue, covering both papers published...

Calculation of resonance integral for fuel cluster

International Nuclear Information System (INIS)

Remsak, S.

1969-01-01

The procedure for calculating the shielding correction, formulated in the previous paper [6], was broadened and applied for a cluster of cylindrical rods. The sam analytical method as in the previous paper was applied. A combination of Gauss method with the method of Almgren and Porn used for solving the same type of integral was used to calculate the geometry functions. CLUSTER code was written for ZUSE-Z-23 computer to calculate the shielding corrections for pairs of fuel rods in the cluster. Computing time for one pair of fuel rods depends on the number of closely placed rod, and for two closely placed rods it is about 3 hours. Calculations were done for clusters containing 7 and 19 UO 2 rods. results show that calculated values of resonance integrals are somewhat higher than the values obtained by Helstrand empirical formula. Taking into account the results for two rods from the previous paper it can be noted that the calculated and empirical values for clusters with 2 and 7 rods are in agreement since the deviations do not exceed the limits of experimental error (±2%). In case of larger cluster with 19 rods deviations are higher than the experimental error. Most probably the calculated values exceed the experimental ones result from the fact that in this paper the shielding correction is calculated only in the region up to 1 keV [sr

Preliminary Investigation on the Behavior of Pore Air Pressure During Rainfall Infiltration

Science.gov (United States)

Ashraf Mohamad Ismail, Mohd; Min, Ng Soon; Hasliza Hamzah, Nur; Hazreek Zainal Abidin, Mohd; Madun, Aziman; Tajudin, Saiful Azhar Ahmad

2018-04-01

This paper focused on the preliminary investigation of pore air pressure behaviour during rainfall infiltration in order to substantiate the mechanism of rainfall induced slope failure. The actual behaviour or pore air pressure during infiltration is yet to be clearly understood as it is regularly assumed as atmospheric. Numerical modelling of one dimensional (1D) soil column was utilized in this study to provide a preliminary insight of this highlighted uncertainty. Parametric study was performed by using rainfall intensities of 1.85 x 10-3m/s and 1.16 x 10-4m/s applied on glass beads to simulate intense and modest rainfall conditions. Analysis results show that the high rainfall intensity causes more development of pore air pressure compared to low rainfall intensity. This is because at high rainfall intensity, the rainwater cannot replace the pore air smoothly thus confining the pore air. Therefore, the effect of pore air pressure has to be taken into consideration particularly during heavy rainfall.

Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

International Nuclear Information System (INIS)

Ma Xubo; Chen Yixue; Gao Bin

2013-01-01

The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

Experience with PET FDG - Preliminary analysis

International Nuclear Information System (INIS)

Massardo, Teresa; Jofre, Josefina; Canessa, Jose; Gonzalez, Patricio; Humeres, Pamela; Sierralta, Paulina; Galaz, Rodrigo; Miranda, Karina

2004-01-01

Full text: The objective of this preliminary communication was to analyse the indications and data in initial group of patients studied with first dedicated PET scanner in the country at Hospital Militar in Santiago Chile. The main application of positron emission tomography (PET) with 18-Fluoro deoxyglucose (FDG) is related with oncological patients management. We studied 136 patients, 131 (97%) with known or suspected malignant disease and remaining 5 for cardiological or neuropsychiatric disease. Ten patients were controlled diabetics (1 insulin dependent). Their mean age was 51.6±18 years ranging from 6 to 84 years and 65% were females. A total of 177 scans were acquired using a dedicated PET (Siemens HR + with 4mm resolution) system. Mean F18-FDG injected dose was 477±107 MBq (12.9±2.9 mCi). Mean blood glucose levels, performed prior the injection, were 94±17mg/dl (range 62-161). F18-FDG was obtained from the cyclotron IBA Cyclone 18/9 installed in the Chilean Agency of Nuclear Energy, distant about 15 miles away from the clinical PET facility. PET studies were analyzed by at least 4 independent observers visually. Standardized uptake value (SUV) was calculated in some cases. Image fusion of FDG images with recent anatomical (CT, MRI) studies was performed where available. Data acquisition protocol consisted in 7-8 beds/study from head to mid-thighs, with 6-7-min/bed acquisitions, 36% transmission with germanium 68 rods. Data was reconstructed with standard OSEM protocol. The main indications included pulmonary lesions in 31%, gastrointestinal cancers in 21%, melanoma in 13% and lymphoma in 9% patients. The remaining were of breast, thyroid, testes, ovary, musculoskeletal (soft tissue and bone), brain tumour etc. Abnormal focal tracer uptake was observed in 83/131 oncological patients, 54% corroborating with clinical diagnosis of primary tumor or recurrence while 46% showed new metastatic localization. FDG scans were normal 36/131 patients. In 9 patients

Experience with PET FDG - Preliminary analysis

Energy Technology Data Exchange (ETDEWEB)

Massardo, Teresa; Jofre, Josefina; Canessa, Jose; Gonzalez, Patricio; Humeres, Pamela; Sierralta, Paulina; Galaz, Rodrigo; Miranda, Karina [Centro PET de Imagenes Moleculares, Hospital Militar de Santiago, Santiago (Chile)

2004-01-01

Full text: The objective of this preliminary communication was to analyse the indications and data in initial group of patients studied with first dedicated PET scanner in the country at Hospital Militar in Santiago Chile. The main application of positron emission tomography (PET) with 18-Fluoro deoxyglucose (FDG) is related with oncological patients management. We studied 136 patients, 131 (97%) with known or suspected malignant disease and remaining 5 for cardiological or neuropsychiatric disease. Ten patients were controlled diabetics (1 insulin dependent). Their mean age was 51.6{+-}18 years ranging from 6 to 84 years and 65% were females. A total of 177 scans were acquired using a dedicated PET (Siemens HR + with 4mm resolution) system. Mean F18-FDG injected dose was 477{+-}107 MBq (12.9{+-}2.9 mCi). Mean blood glucose levels, performed prior the injection, were 94{+-}17mg/dl (range 62-161). F18-FDG was obtained from the cyclotron IBA Cyclone 18/9 installed in the Chilean Agency of Nuclear Energy, distant about 15 miles away from the clinical PET facility. PET studies were analyzed by at least 4 independent observers visually. Standardized uptake value (SUV) was calculated in some cases. Image fusion of FDG images with recent anatomical (CT, MRI) studies was performed where available. Data acquisition protocol consisted in 7-8 beds/study from head to mid-thighs, with 6-7-min/bed acquisitions, 36% transmission with germanium 68 rods. Data was reconstructed with standard OSEM protocol. The main indications included pulmonary lesions in 31%, gastrointestinal cancers in 21%, melanoma in 13% and lymphoma in 9% patients. The remaining were of breast, thyroid, testes, ovary, musculoskeletal (soft tissue and bone), brain tumour etc. Abnormal focal tracer uptake was observed in 83/131 oncological patients, 54% corroborating with clinical diagnosis of primary tumor or recurrence while 46% showed new metastatic localization. FDG scans were normal 36/131 patients. In 9

Carbon Footprint Calculator | Climate Change | US EPA

Science.gov (United States)

2016-12-12

An interactive calculator to estimate your household's carbon footprint. This tool will estimate carbon pollution emissions from your daily activities and show how to reduce your emissions and save money through simple steps.

The MCEF code for nuclear evaporation and fission calculations

International Nuclear Information System (INIS)

Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J.; Arruda-Neto, J.D.T.; Rodriguez, O.; Goncalves, M.

2001-11-01

We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)

Development of hybrid core calculation system using 2-D full-core heterogeneous transport calculation and 3-D advanced nodal calculation

International Nuclear Information System (INIS)

Sugimura, Naoki; Mori, Masaaki; Hijiya, Masayuki; Ushio, Tadashi; Arakawa, Yasushi

2004-01-01

This paper presents the Hybrid Core Calculation System which is a very rigorous but a practical calculation system applicable to best estimate core design calculations taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D core calculation results under the consistent calculation condition and then apply them for SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step with the core conditions previously determined by ordinary SIMULATE-3 core calculation to avoid time consuming iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. The final SIMULATE-3 3-D calculation using the correction factors is performed with iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. (author)

Relationship Dissolution and Psychologically Aggressive Dating Relationships: Preliminary Findings From a College-Based Relationship Education Course.

Science.gov (United States)

Negash, Sesen; Cravens, Jaclyn D; Brown, Preston C; Fincham, Frank D

This study evaluated the impact of a relationship education program, delivered as part of a college course, among students (N = 152) who reported experiencing psychological aggression in their exclusive dating relationship. Preliminary results showed that compared to those in the control group, participants receiving relationship education were significantly more likely to end their romantic relationship, even after controlling for relationship satisfaction. Furthermore, when relationship termination occurred, those in the intervention group were significantly more likely to attribute the breakup to their participation in the class as compared to those in the control group. The tentative findings are an important preliminary step in assessing the benefits of relationship education in reducing the risk of psychological aggression among college students.

Sensitivity Measurement of Transmission Computer Tomography: thePreliminary Experimental Study

International Nuclear Information System (INIS)

Widodo, Chomsin-S; Sudjatmoko; Kusminarto; Agung-BS Utomo; Suparta, Gede B

2000-01-01

This paper reports result of preliminary experimental study onmeasurement method for sensitivity of a computed tomography (CT) scanner. ACT scanner has been build at the Department of Physics, FMIPA UGM and itsperformance based on its sensitivity was measured. The result showed that themeasurement method for sensitivity confirmed this method may be developedfurther as a measurement standard. Although the CT scanner developed has anumber of shortcoming, the analytical results from the sensitivitymeasurement suggest a number of reparations and improvements for the systemso that improved reconstructed CT images can be obtained. (author)

SPANISH MULTICENTRIC STUDY ABOUT NUTRITION-INFLAMATIONhn WITH MID DILUTION (ENIMID STUDY: PRELIMINARY RESULTS

Directory of Open Access Journals (Sweden)

Barril G

2012-06-01

CONCLUSIONS: 1-The preliminary results show that MidDilution provides a good removal of small and middle molecules, increases appetite by providing a proper balance of cytokines through stimulation of antiinflamatory ones and neuropeptide Y. 2-It provides an improvement of body composition. Finally MidDilution improves nutritional parameters which leads to a better quality of life, as well as physical and mental status.

Advances in supercell calculation methods and comparison with measurements

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Baril, R; Hotte, G [Hydro-Quebec, Central Nucleaire Gentilly, Montreal, Quebec (Canada)

1996-07-01

In the last few years, modelling techniques have been developed in new supercell computer codes. These techniques have been used to model the CANDU reactivity devices. One technique is based on one- and two-dimensional transport calculations with the WIMS-AECL lattice code followed by super homogenization and three-dimensional flux calculations in a modified version of the MULTICELL code. The second technique is based on two- and three-dimensional transport calculations in DRAGON. The code calculates the lattice properties by solving the transport equation in a two-dimensional geometry followed by supercell calculations in three dimensions. These two calculation schemes have been used to calculate the incremental macroscopic properties of CANDU reactivity devices. The supercell size has also been modified to define incremental properties over a larger region. The results show improved agreement between the reactivity worth of zone controllers and adjusters. However, at the same time the agreement between measured and simulated flux distributions deteriorated somewhat. (author)

Innovative analytical competence. Optimization of shielding components and lifetime activation calculations

Energy Technology Data Exchange (ETDEWEB)

Boehlke, Steffen; Wortmann, Birgit; Aguilar, Arturo Lizon [STEAG Energy Services GmbH, Essen (Germany)

2014-08-15

Shielding and activation calculations always require a high level of engineering competence and powerful hard- and software tools. With the application of current methods often certain limits were reached in the past. The engineering work for optimization efforts regarding complex components with high shielding requirements exceeded the savings in material. With regard to activation the challenges in size of the geometric model and considered operation time rises constantly and pushes computing time beyond reasonable time frames. These challenges require the application of new and faster methodologies. The application of new and innovative methods is presented for a shielding optimization project to decrease the radiation level, to keep the dose rate limits, and to reduce the amount of used shielding material. In a second case a prediction of the activated materials with it's dose distribution in the surrounding area and classification of waste quantities in the structural materials of a nuclear reactor is presented. For the shielding project the preliminary design CAD model was imported into the software tool, several iterations were run and a significantly reduced radiation exposure together with a significant reduction in shieling material were achieved. For the activation calculations it could be demonstrated that it is possible to determine the activation, waste quantities and dose distribution for the structural materials of a nuclear reactor based on lifetime operational data within reasonable time frames.

Calculations of air cooler for new subsonic wind tunnel

Science.gov (United States)

Rtishcheva, A. S.

2017-10-01

As part of the component development of TsAGI’s new subsonic wind tunnel where the air flow velocity in the closed test section is up to 160 m/sec hydraulic and thermal characteristics of air cooler are calculated. The air cooler is one of the most important components due to its highest hydraulic resistance in the whole wind tunnel design. It is important to minimize its hydraulic resistance to ensure the energy efficiency of wind tunnel fans and the cost-cutting of tests. On the other hand the air cooler is to assure the efficient cooling of air flow in such a manner as to maintain the temperature below 40 °C for seamless operation of measuring equipment. Therefore the relevance of this project is driven by the need to develop the air cooler that would demonstrate low hydraulic resistance of air and high thermal effectiveness of heat exchanging surfaces; insofar as the cooling section must be given up per unit time with the amount of heat Q=30 MW according to preliminary evaluations. On basis of calculation research some variants of air cooler designs are proposed including elliptical tubes, round tubes, and lateral plate-like fins. These designs differ by the number of tubes and plates, geometrical characteristics and the material of finned surfaces (aluminium or cooper). Due to the choice of component configurations a high thermal effectiveness is achieved for finned surfaces. The obtained results form the basis of R&D support in designing the new subsonic wind tunnel.

Smartphone-based fluorescence spectroscopy device aiding in preliminary skin screening

Science.gov (United States)

Sahoo, Aparajita; Wahi, Akshat; Das, Anshuman

2018-02-01

Preliminary diagnosis of closely resembling skin conditions can be highly subjective for dermatologists. In ambiguous cases, it often leads to performing invasive procedures like biopsies. Different skin conditions, however, have varying concentrations of fluorophores (like collagen, NADH) and chromophores (like melanin, hemoglobin) which can alter their fluorescence spectra. We demonstrate a handheld, portable, smartphone-based spectrometer that leverages these alterations in skin autofluorescence spectra for rapid screening of skin conditions. This methodology involves excitation of affected skin areas with ultraviolet (UV-A) 385 nm light, capturing the generated fluorescence spectra and sending the data wirelessly to a companion mobile application for data storage, analysis and visualization. By collecting the fluorescence spectral signals from healthy and unhealthy skin conditions, we establish that the signals collected using this portable device can be used to develop a classification method to help in differentially diagnosing these conditions. It shows promise as a useful skin screening tool for both dermatologists and primary health care workers. This device can enable quick, non-invasive and a more objective preliminary examination. We envision the device to be especially useful in primary healthcare centers of developing countries where availability of dermatologists is limited.

Reference Monte Carlo calculations of Maria reactor core

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

2002-01-01

The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

Kemungkinan Penerapan Preliminary Ruling Procedure sebagai Media Constitutional Complaint di Mahkamah Konstitusi

OpenAIRE

Arundhati, Gautama Budi

2017-01-01

Preliminary ruling procedure seperti yang diterapkan di Uni Eropa dapat menjadi metode alternatif dalam pelaksanaan constitutional complaint di Indonesia. Undang-Undang Dasar 1945 sebagai hukum tertinggi di Indonesia dikawal oleh lembaga yang bernama Mahkamah Konstitusi Republik Indonesia, dan dalam preliminary ruling procedure untuk pemberlakuan constitutional complaint maka dibutuhkan Peran Pengadilan Negeri dimana melalui preliminary ruling procedure tersebut dapat melakukan constitutional...