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Sample records for method mcnp4c code

  1. A group of neutronics calculations in the MNSR using the MCNP-4C code

    International Nuclear Information System (INIS)

    Khattab, K.; Sulieman, I.

    2009-11-01

    The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)

  2. Nuclear densimeter of soil simulated in MCNP-4C code

    International Nuclear Information System (INIS)

    Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.

    2009-01-01

    The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)

  3. Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

    2010-01-01

    The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

  4. Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C

    International Nuclear Information System (INIS)

    Suman, H.; Kharita, M. H.; Yousef, S.

    2008-02-01

    In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)

  5. Calibration of a foot borne spectrometry system using the MCNP 4C code

    International Nuclear Information System (INIS)

    Nylen, T.; Agren, G.

    2004-01-01

    The increased interest for the cycling of radioactive Caesium in natural ecosystems has gained need for rapid and reliable methods to investigate the deposition density in natural soils. One commonly used method, soil sampling, is a good method that correctly used gives information of both the horizontal and vertical distribution of the desired nuclide. The main disadvantage is that the method is time consuming regarding sampling, preparation and measurements. An alternative method is the use of semiconductors or scintillation detectors in the field i.e. in cars, airplanes, or helicopters. Theses methods are rapid and integrate over large areas which gives a more reliable mean value provided that the operator has some basic knowledge about the depth distribution of the radio nuclides and bulk density in the soil. To be effective the systems are often connected to a GPS to give the exact coordinate for each measurement. In a situation where the area of interest is too large to cover by soil samples and measurements by airplane not will give a spatial resolution good enough, one feasible method is to use a foot borne gamma spectrometry system. The advantage of a foot borne system is that the operator can cover a quite large area within a few hours and that the method can detect small anomalies in the deposition field which may be difficult to discover with soil samples. This abstract describes the calibration of a foot borne gamma-spectrometry system carried in a back-pack and consisting of a NaI-detector, a GPS and a system for logging activity and position. The detector system and surroundings has been modeled in the Monte Carlo code MCNP 4C (Figure 1). The Monte Carlo method gives the possibility to study the influence of complex geometries that are difficult to create for a practical calibration using real activity. The results of the MCNP calibration model, has been compared to foot borne gamma-spectrometry field measurements in a Cs-137 deposition area. A

  6. A simulation of a pebble bed reactor core by the MCNP-4C computer code

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    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  7. Performance of the MTR core with MOX fuel using the MCNP4C2 code

    International Nuclear Information System (INIS)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-01-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

  8. GB - a preliminary linking code between MCNP4C and Origen2.1 - DEN/UFMG version

    International Nuclear Information System (INIS)

    Campolina, Daniel; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Cavatoni, Andre

    2009-01-01

    Nowadays it is possible to perform burnup simulation in a detailed 3D geometry and a continuous energy description by the Monte Carlo method. This paper describes an initial project to create and verify a connection code to link Origen2.1 (Oak Ridge National Laboratory) and MCNP4C (Los Alamos National Laboratory). Essentially the code includes point depletion capability to the MCNP code. The incorporation of point depletion capability is explicit and can be summarized by three steps: 1-Monte Carlo determines reaction rates, 2-the reaction rates are used to determine microscopic cross sections for depletion equations, 3-solution of depletion equations (given by Origen2.1) determines number densities for next MCNP step. To evaluate the initial version of the program, we focused on comparing the results with one of the major Monte Carlo burnup codes: MCNPX version 2.6.0. The input files for all codes share the same MCNP geometry, nuclear data library and core thermal power. While simulating 75 time steps at 800 kw of a Heat Pipe Power System model, we have found that the codes generate very similar results. The neutron flux and criticality value of the core agree, especially in the begin of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB (author)

  9. Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    Dawahra, S.; Khattab, K.

    2011-01-01

    Highlights: → The MCNP4C code was used to calculate the power distribution in 3-D geometry in the MNSR reactor. → The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. → The minimum power was found in the fuel ring number 9 and was 79.9 W. → The total power in the total fuel rods was 30.9 kW. - Abstract: The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.

  10. S values at voxels level for 188Re and 90Y calculated with the MCNP-4C code

    International Nuclear Information System (INIS)

    Coca Perez, Marco Antonio; Torres Aroche, Leonel Alberto; Cornejo, Nestor; Martin Hernandez, Guido

    2003-01-01

    The main objective of this work was estimate the voxels S values for 188 Re at cubical geometry using the MCNP-4C code for the simulation of radiation transport and energy deposition. Mean absorbed dose to target voxels per radioactive decay in a source voxels were estimated and reported for 188 Re and Y 90 . A comparison of voxels S values computed with the MCNP code the data reported in MIRD pamphlet 17 for 90 Y was performed in order to evaluate our results

  11. Monte Carlo simulation for treatment planning optimization of the COMS and USC eye plaques using the MCNP4C code

    International Nuclear Information System (INIS)

    Jannati Isfahani, A.; Shokrani, P.; Raisali, Gh.

    2010-01-01

    Ophthalmic plaque radiotherapy using I-125 radioactive seeds in removable episcleral plaques is often used in management of ophthalmic tumors. Radioactive seeds are fixed in a gold bowl-shaped plaque and the plaque is sutured to the scleral surface corresponding to the base of the intraocular tumor. This treatment allows for a localized radiation dose delivery to the tumor with a minimum target dose of 85 Gy. The goal of this study was to develop a Monte Carlo simulation method for treatment planning optimization of the COMS and USC eye plaques. Material and Methods: The MCNP4C code was used to simulate three plaques: COMS-12mm, COMS-20mm, and USC ≠9 with I-125 seeds. Calculation of dose was performed in a spherical water phantom (radius 12 mm) using a 3D matrix with a size of 12 voxels in each dimension. Each voxel contained a sphere of radius 1 mm. Results: Dose profiles were calculated for each plaque. Isodose lines were created in 2 planes normal to the axes of the plaque, at the base of the tumor and at the level of the 85 Gy isodose in a 7 day treatment. Discussion and Conclusion: This study shows that it is necessary to consider the following tumor properties in design or selection of an eye plaque: the diameter of tumor base, its thickness and geometric shape, and the tumor location with respect to normal critical structures. The plaque diameter is selected by considering the tumor diameter. Tumor thickness is considered when selecting the seed parameters such as their number, activity and distribution. Finally, tumor shape and its location control the design of following parameters: the shape and material of the plaque and the need for collimation.

  12. Investigation of Anisotropy Caused by Cylinder Applicator on Dose Distribution around Cs-137 Brachytherapy Source using MCNP4C Code

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    Sedigheh Sina

    2011-06-01

    Full Text Available Introduction: Brachytherapy is a type of radiotherapy in which radioactive sources are used in proximity of tumors normally for treatment of malignancies in the head, prostate and cervix. Materials and Methods: The Cs-137 Selectron source is a low-dose-rate (LDR brachytherapy source used in a remote afterloading system for treatment of different cancers. This system uses active and inactive spherical sources of 2.5 mm diameter, which can be used in different configurations inside the applicator to obtain different dose distributions. In this study, first the dose distribution at different distances from the source was obtained around a single pellet inside the applicator in a water phantom using the MCNP4C Monte Carlo code. The simulations were then repeated for six active pellets in the applicator and for six point sources.  Results: The anisotropy of dose distribution due to the presence of the applicator was obtained by division of dose at each distance and angle to the dose at the same distance and angle of 90 degrees. According to the results, the doses decreased towards the applicator tips. For example, for points at the distances of 5 and 7 cm from the source and angle of 165 degrees, such discrepancies reached 5.8% and 5.1%, respectively.  By increasing the number of pellets to six, these values reached 30% for the angle of 5 degrees. Discussion and Conclusion: The results indicate that the presence of the applicator causes a significant dose decrease at the tip of the applicator compared with the dose in the transverse plane. However, the treatment planning systems consider an isotropic dose distribution around the source and this causes significant errors in treatment planning, which are not negligible, especially for a large number of sources inside the applicator.

  13. S values at voxels level for 188Re and 90Y calculated with the MCNP-4C code

    International Nuclear Information System (INIS)

    Coca, M.A.; Torres, L.A.; Cornejo, N.; Martin, G.

    2008-01-01

    Full text: MIRD formalism at voxel level has been suggested as an optional methodology to perform internal radiation dosimetry calculation during internal radiation therapy in Nuclear Medicine. Voxel S values for Y 90 , 131 I, 32 P, 99m Tc and 89 Sr have been published to different sizes. Currently, 188 Re has been proposed as a promising radionuclide for therapy due to its physical features and availability from generators. The main objective of this work was to estimate the voxel S values for 188 Re at cubical geometry using the MCNP-4C code for the simulations of radiation transport and energy deposition. Mean absorbed dose to target voxels per radioactive decay in a source voxel were estimated and reported for 188 Re and Y 90 . A comparison of voxel S values computed with the MCNP code and the data reported in MIRD Pamphlet 17 for 90 Y was performed in order to evaluate our results. (author)

  14. Evaluation of dose equivalent to the people accompanying patients in diagnostic radiology using MCNP4C Monte Carlo code

    International Nuclear Information System (INIS)

    Mehdizadeh, S.; Faghihi, R.; Sina, S.; Zehtabian, M.

    2007-01-01

    Complete text of publication follows. Objective: X rays used in diagnostic radiology contribute a major share to population doses from man-made sources of radiation. In some branches of radiology, it is necessary that another person stay in the imaging room and immobilize the patient to carry out radiological operation. ICRP 70 recommends that this should be done by parents or accompanying nursing or ancillary personnel and not in any case by radiation workers. Methods: Dose measurements were made previously using standard methods employing LiF TLD-100 dosimeters. A TLD card was installed on the main trunk of the body of the accompanying people where the maximum dose was probable. In this research the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) is used to calculate the equivalent dose to the people accompanying patients exposed to radiation scattered from the patient (Without protective clothing). To do the simulations, all components of the geometry are placed within an air-filled box. Two homogeneous water phantoms are used to simulate the patient and the accompanying person. The accompanying person leans against the table at one side of the patient. Finally in case of source specification, only the focus of the X-ray tube is modelled, i.e. as a standard MCNP point source emitting a cone of photons. Photon stopping material is used as a collimator model to reduce the circular cross section of the cone to a rectangle. The X-ray spectra to be used in the MCNP simulations are generated with spectrum generator software, taking the X-ray voltage and all filtration applied in the clinic as input parameters. These calculations are done for different patient sizes and for different radiological operations. Results: In case of TL dosimetry, for a group of 100 examinations, the dose equivalents ranged from 0.01 μsv to 0.13 msv with the average of 0.05 msv. The results are seen to be in close agreement with Monte Carlo simulations

  15. Gamma ray shielding study of barium-bismuth-borosilicate glasses as transparent shielding materials using MCNP-4C code, XCOM program, and available experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

    2017-02-15

    In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

  16. Gamma Ray Shielding Study of Barium–Bismuth–Borosilicate Glasses as Transparent Shielding Materials using MCNP-4C Code, XCOM Program, and Available Experimental Data

    Directory of Open Access Journals (Sweden)

    Reza Bagheri

    2017-02-01

    Full Text Available In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium–bismuth–borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium–bismuth–borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

  17. Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    Dawahra, S.; Khattab, K.

    2012-01-01

    The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 k W. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 k W. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively. (author)

  18. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

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    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  19. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  20. Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR

    International Nuclear Information System (INIS)

    Shaaban, I.

    2009-11-01

    Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)

  1. Dosimetry analysis of distributions radials dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radias de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, Talita S.; Yoriyaz, Helio [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, Marco A.R., E-mail: tasallesc@gmail.co [UNESP, Botucatu, SP (Brazil). Faculdade de Medicina. Servico de Radioterapia; Louzada, Mario J.Q. [UNESP, Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

    2011-07-01

    Although they are no longer manufactured, the applicators of {sup 90}Sr + {sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been re calibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr + {sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radio chromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radio chromium films for this type of dosimetry. (author)

  2. Dosimetry analysis of distribution radial dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radiais de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, T.S.; Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, M.A.R. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Servico de Radioterapia; Louzada, M.J.Q. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

    2010-07-01

    Although they are no longer manufactured, the applicators of {sup 90}Sr +{sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been recalibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr+{sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radiochromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radiochromium films for this type of dosimetry. (author)

  3. Gamma knife simulation using the MCNP4C code and the zubal phantom and comparison with experimental data

    International Nuclear Information System (INIS)

    Gholami, S.; Kamali Asl, A.; Aghamiri, M.; Allahverdi, M.

    2010-01-01

    Gamma Knife is an instrument specially designed for treating brain disorders. In Gamma Knife, there are 201 narrow beams of cobalt-60 sources that intersect at an isocenter point to treat brain tumors. The tumor is placed at the isocenter and is treated by the emitted gamma rays. Therefore, there is a high dose at this point and a low dose is delivered to the normal tissue surrounding the tumor. Material and Method: In the current work, the MCNP simulation code was used to simulate the Gamma Knife. The calculated values were compared to the experimental ones and previous works. Dose distribution was compared for different collimators in a water phantom and the Zubal brain-equivalent phantom. The dose profiles were obtained along the x, y and z axes. Result: The evaluation of the developed code was performed using experimental data and we found a good agreement between our simulation and experimental data. Discussion: Our results showed that the skull bone has a high contribution to both scatter and absorbed dose. In other words, inserting the exact material of brain and other organs of the head in digital phantom improves the quality of treatment planning. This work is regarding the measurement of absorbed dose and improving the treatment planning procedure in Gamma-Knife radiosurgery in the brain.

  4. Gamma Knife Simulation Using the MCNP4C Code and the Zubal Phantom and Comparison with Experimental Data

    Directory of Open Access Journals (Sweden)

    Somayeh Gholami

    2010-06-01

    Full Text Available Introduction: Gamma Knife is an instrument specially designed for treating brain disorders. In Gamma Knife, there are 201 narrow beams of cobalt-60 sources that intersect at an isocenter point to treat brain tumors. The tumor is placed at the isocenter and is treated by the emitted gamma rays. Therefore, there is a high dose at this point and a low dose is delivered to the normal tissue surrounding the tumor. Material and Method: In the current work, the MCNP simulation code was used to simulate the Gamma Knife. The calculated values were compared to the experimental ones and previous works. Dose distribution was compared for different collimators in a water phantom and the Zubal brain-equivalent phantom. The dose profiles were obtained along the x, y and z axes. Result: The evaluation of the developed code was performed using experimental data and we found a good agreement between our simulation and experimental data. Discussion: Our results showed that the skull bone has a high contribution to both scatter and absorbed dose. In other words, inserting the exact material of brain and other organs of the head in digital phantom improves the quality of treatment planning. This work is regarding the measurement of absorbed dose and improving the treatment planning procedure in Gamma-Knife radiosurgery in the brain.

  5. Investigation of the Effects of Tissue Inhomogeneities on the Dosimetric Parameters of a Cs-137 Brachytherapy Source using the MCNP4C Code

    Directory of Open Access Journals (Sweden)

    Mehdi Zehtabian

    2010-09-01

    Full Text Available Introduction: Brachytherapy is the use of small encapsulated radioactive sources in close vicinity of tumors. Various methods are used to obtain the dose distribution around brachytherapy sources. TG-43 is a dosimetry protocol proposed by the AAPM for determining dose distributions around brachytherapy sources. The goal of this study is to update this protocol for presence of bone and air inhomogenities.  Material and Methods: To update the dose rate constant parameter of the TG-43 formalism, the MCNP4C simulations were performed in phantoms composed of water-bone and water-air combinations. The values of dose at different distances from the source in both homogeneous and inhomogeneous phantoms were estimated in spherical tally cells of 0.5 mm radius using the F6 tally. Results: The percentages of dose reductions in presence of air and bone inhomogenities for the Cs-137 source were found to be 4% and 10%, respectively. Therefore, the updated dose rate constant (Λ will also decrease by the same percentages.   Discussion and Conclusion: It can be easily concluded that such dose variations are more noticeable when using lower energy sources such as Pd-103 or I-125.

  6. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2014-12-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  7. Calculation of radiation dose rate above water layer of Interim Spent Fuel Storage Jaslovske Bohunice by the point Kernels (VISIPLAN) and Monte Carlo (MCNP4C) methods

    International Nuclear Information System (INIS)

    Slavik, O.; Kucharova, D.; Listjak, M.; Fueloep, M.

    2008-01-01

    The aim of this paper is to evaluate maximal dose rate (DR) of gamma radiation above different configurations of reservoirs with spent nuclear fuel with cooling period 1.8 year and to compare by buildup factor method (Visiplan) and Monte Carlo simulations and to appreciate influence of scattered photons in the case of calculation of fully filled fuel transfer storage (FTS). On the ground of performed accounts it was shown, that relative contributions of photons from adjacent reservoirs are in the case buildup factor method (Visiplan) similar to Monte Carlo simulations. It means, that Visiplan can be used also for valuation of contributions of of dose rates from neighbouring reservoirs. It was shown, that calculations of DR by Visiplan are conservatively overestimated for this source of radiation and thickness of shielding approximately 2.6 - 3 times. Also following these calculations resulted, that by storage of reservoirs with cooling period 1.8 years in FTS is not needed any additional protection measures for workers against primal safety report. Calculated DR also above fully filled FTS by these reservoirs in Jaslovske Bohunice is very low on the level 0.03 μSv/h. (authors)

  8. Calculation of radiation dose rate above water layer of Interim Spent Fuel Storage Jaslovske Bohunice by the point Kernels (VISIPLAN) and Monte Carlo (MCNP4C) methods

    International Nuclear Information System (INIS)

    Slavik, O.; Kucharova, D.; Listjak, M.; Fueloep, M.

    2009-01-01

    The aim of this paper is to evaluate maximal dose rate (DR) of gamma radiation above different configurations of reservoirs with spent nuclear fuel with cooling period 1.8 year and to compare by buildup factor method (Visiplan) and Monte Carlo simulations and to appreciate influence of scattered photons in the case of calculation of fully filled fuel transfer storage (FTS). On the ground of performed accounts it was shown, that relative contributions of photons from adjacent reservoirs are in the case buildup factor method (Visiplan) similar to Monte Carlo simulations. It means, that Visiplan can be used also for valuation of contributions of of dose rates from neighbouring reservoirs. It was shown, that calculations of DR by Visiplan are conservatively overestimated for this source of radiation and thickness of shielding approximately 2.6 - 3 times. Also following these calculations resulted, that by storage of reservoirs with cooling period 1.8 years in FTS is not needed any additional protection measures for workers against primal safety report. Calculated DR also above fully filled FTS by these reservoirs in Jaslovske Bohunice is very low on the level 0.03 μSv/h. (authors)

  9. Monte Carlo simulation applied to radiosurgery narrow beams using MCNP-4C

    International Nuclear Information System (INIS)

    Chaves, A.; Lopes, M.C.; Oliveira, C.

    2001-01-01

    Dose measurements for the narrow photon beams used in radiosurgery are complicated by the lack of electron equilibrium which is a requirement namely for ionometric methods. To overcome this difficulty the use of different dosimetric supports is strongly recommended in order to appreciate the influence of each type of detector. Monte Carlo simulation is another kind of tool to assess the details of the energy deposition phenomena in such narrow photon beams. In this study output factors and depth dose calculated by the Monte Carlo MCNP-4C code are presented and compared with experimental data measured with a diode, a Markus chamber, a 0.125 cc thimble chamber and a Pinpoint chamber. Simulated energy spectra for narrow beams are also presented in order to compare them with the reference 10 cm x 10 cm beam field size and thus discuss the different contributions of the absorbed energy in water, in each case. A detailed analysis on the photon energy spectra showed a slight decrease on the photon mean energy that can be explained by the increased scattering inside the additional collimators. Calculated and measured depth doses curves are in good agreement for most of the collimators. For the two smallest collimators some differences have been pointed and explained according to the characteristics of the detectors (author)

  10. MCNP4C2, Coupled Neutron, Electron Gamma 3-D Time-Dependent Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. MCNP4C2 is an interim release of MCNP4C with distribution restricted to the Criticality Safety community and attendees of the LANL MCNP workshops. The major new features of MCNP4C2 include: - Photonuclear physics; - Interactive plotting; - Plot superimposed weight window mesh; - Implement remaining macro-body surfaces; - Upgrade macro-bodies to surface sources and other capabilities; - Revised summary tables; - Weight window improvements. See the MCNP home page more information http://www-xdiv.lanl.gov/XCI/PROJECTS/MCNP with a link to the MCNP Forum. See the Electronic Notebook at http://www-rsicc.ornl.gov/rsic.html for information on user experiences with MCNP. 2 - Methods:MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. Pointwise continuous-energy cross section data are used, although multigroup data may also be used. Fixed-source adjoint calculations may be made with the multigroup data option. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to

  11. DXRaySMCS. First user friendly interface developed for prediction of diagnostic radiology X-ray spectra produced by Monte Carlo (MCNP-4C) simulation in Iran

    International Nuclear Information System (INIS)

    Bahreyni Toossi, M.T.; Zare, H.; Moradi Faradanbe, H.

    2008-01-01

    An accurate knowledge of the output energy spectra of an x-ray tube is essential in many areas of radiological studies. It forms the basis of almost all image quality simulations and enable system designers to predict patient dose more accurately. Many radiological physics problems that can be solved by Monte Carlo simulation methods require an x-ray spectra as input data. Computer simulation of x-ray spectra is one of the most important tools for investigation of patient dose and image quality in diagnostic radiology systems. In this work the general purpose Monte Carlo N-particle radiation transport computer code (MCNP-4C) was used for the simulation of x-ray spectra in diagnostic radiology, Electron's path in the target was followed until it's energy was reduced to 10 keV. A user friendly interface named 'Diagnostic X-ray Spectra by Monte Carlo Simulation (DXRaySMCS)' was developed to facilitate the application of MCNP-4C code for diagnostic radiology spectrum prediction. The program provides a user friendly interface for modifying the MCNP input file, launching the MCNP program to simulate electron and photon transport and processing the MCNP output file to yield a summary of the results (Relative Photon Number per Energy Bin). In this article the development and characteristics of DXRaySMCS are outlined. As part of the validation process, out put spectra for 46 diagnostic radiology system settings produced by DXRaySMCS were compared with the corresponding IPEM78. Generally, there is a good agreement between the two sets of spectra. No statistically significant differences have been observed between IPEM78 reported spectra and the simulated spectra generated in this study. (author)

  12. Monte Carlo simulation of x-ray spectra in diagnostic radiology and mammography using MCNP4C

    Energy Technology Data Exchange (ETDEWEB)

    Ay, M R [Department of Physics and Nuclear Sciences, AmirKabir University of Technology, Tehran (Iran, Islamic Republic of); Shahriari, M [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Sarkar, S [Department of Medical Physics, Tehran University of Medical Science, Tehran (Iran, Islamic Republic of); Adib, M [TPP Co., GE Medical Systems, Iran Authorized Distributor, Tehran (Iran, Islamic Republic of); Zaidi, H [Division of Nuclear Medicine, Geneva University Hospital, 1211 Geneva (Switzerland)

    2004-11-07

    The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

  13. Monte Carlo simulation of x-ray spectra in diagnostic radiology and mammography using MCNP4C

    Science.gov (United States)

    Ay, M. R.; Shahriari, M.; Sarkar, S.; Adib, M.; Zaidi, H.

    2004-11-01

    The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

  14. A comparison of MCNP4C electron transport with ITS 3.0 and experiment at incident energies between 100 keV and 20 MeV: influence of voxel size, substeps and energy indexing algorithm

    International Nuclear Information System (INIS)

    Schaart, Dennis R.; Jansen, Jan Th.M.; Zoetelief, Johannes; Leege, Piet F.A. de

    2002-01-01

    The condensed-history electron transport algorithms in the Monte Carlo code MCNP4C are derived from ITS 3.0, which is a well-validated code for coupled electron-photon simulations. This, combined with its user-friendliness and versatility, makes MCNP4C a promising code for medical physics applications. Such applications, however, require a high degree of accuracy. In this work, MCNP4C electron depth-dose distributions in water are compared with published ITS 3.0 results. The influences of voxel size, substeps and choice of electron energy indexing algorithm are investigated at incident energies between 100 keV and 20 MeV. Furthermore, previously published dose measurements for seven beta emitters are simulated. Since MCNP4C does not allow tally segmentation with the *F8 energy deposition tally, even a homogeneous phantom must be subdivided in cells to calculate the distribution of dose. The repeated interruption of the electron tracks at the cell boundaries significantly affects the electron transport. An electron track length estimator of absorbed dose is described which allows tally segmentation. In combination with the ITS electron energy indexing algorithm, this estimator appears to reproduce ITS 3.0 and experimental results well. If, however, cell boundaries are used instead of segments, or if the MCNP indexing algorithm is applied, the agreement is considerably worse. (author)

  15. DXRaySMCS: a user-friendly interface developed for prediction of diagnostic radiology X-ray spectra produced by Monte Carlo (MCNP-4C) simulation.

    Science.gov (United States)

    Bahreyni Toossi, M T; Moradi, H; Zare, H

    2008-01-01

    In this work, the general purpose Monte Carlo N-particle radiation transport computer code (MCNP-4C) was used for the simulation of X-ray spectra in diagnostic radiology. The electron's path in the target was followed until its energy was reduced to 10 keV. A user-friendly interface named 'diagnostic X-ray spectra by Monte Carlo simulation (DXRaySMCS)' was developed to facilitate the application of MCNP-4C code for diagnostic radiology spectrum prediction. The program provides a user-friendly interface for: (i) modifying the MCNP input file, (ii) launching the MCNP program to simulate electron and photon transport and (iii) processing the MCNP output file to yield a summary of the results (relative photon number per energy bin). In this article, the development and characteristics of DXRaySMCS are outlined. As part of the validation process, output spectra for 46 diagnostic radiology system settings produced by DXRaySMCS were compared with the corresponding IPEM78. Generally, there is a good agreement between the two sets of spectra. No statistically significant differences have been observed between IPEM78 reported spectra and the simulated spectra generated in this study.

  16. Calculation of the store house worker dose in a lost wax foundry using MCNP-4C

    International Nuclear Information System (INIS)

    Alegria, N.; Legarda, F.; Herranz, M.; Idoeta, R.

    2005-01-01

    Lost wax casting is an industrial process which permits the transmutation into metal of models made in wax. The wax model is covered with a siliceous shell of the required thickness and once this shell is built the set is heated and wax melted. Liquid metal is then cast into the shell replacing the wax. When the metal is cool, the shell is broken away in order to recover the metallic piece. In this process zircon sands are used for the preparation of the siliceous shell. These sands have varying concentrations of natural radionuclides: 238 U, 232 Th and 235 U together with their progenics. The zircon sand is distributed in bags of 50 kg, and 30 bags are on a pallet, weighing 1,500 kg. The pallets with the bags have dimensions 80 cm x 120 cm x 80 cm, and constitute the radiation source in this case. The only pathway of exposure to workers in the store house is external radiation. In this case there is no dust because the bags are closed and covered by plastic, the store house has a good ventilation rate and so radon accumulation is not possible. The workers do not touch with their hands the bags and consequently skin contamination will not take place. In this study all situations of external irradiation to the workers have been considered; transportation of the pallets from vehicle to store house, lifting the pallets to the shelf, resting of the stock on the shelf, getting down the pallets, and carrying the pallets to production area. Using MCNP-4C exposure situations have been simulated, considering that the source has a homogeneous composition, the minimum stock in the store house is constituted by 7 pallets, and the several distances between pallets and workers when they are at work. The photons flux obtained by MCNP-4C is multiplied by the conversion factor of Flux to Kerma for air by conversion factor to Effective Dose by Kerma unit, and by the number of emitted photons. Those conversion factors are obtained of ICRP 74 table 1 and table 17 respectively. This is

  17. Calculation of the store house worker dose in a lost wax foundry using MCNP-4C.

    Science.gov (United States)

    Alegría, Natalia; Legarda, Fernando; Herranz, Margarita; Idoeta, Raquel

    2005-01-01

    Lost wax casting is an industrial process which permits the transmutation into metal of models made in wax. The wax model is covered with a silicaceous shell of the required thickness and once this shell is built the set is heated and wax melted. Liquid metal is then cast into the shell replacing the wax. When the metal is cool, the shell is broken away in order to recover the metallic piece. In this process zircon sands are used for the preparation of the silicaceous shell. These sands have varying concentrations of natural radionuclides: 238U, 232Th and 235U together with their progenics. The zircon sand is distributed in bags of 50 kg, and 30 bags are on a pallet, weighing 1,500 kg. The pallets with the bags have dimensions 80 cm x 120 cm x 80 cm, and constitute the radiation source in this case. The only pathway of exposure to workers in the store house is external radiation. In this case there is no dust because the bags are closed and covered by plastic, the store house has a good ventilation rate and so radon accumulation is not possible. The workers do not touch with their hands the bags and consequently skin contamination will not take place. In this study all situations of external irradiation to the workers have been considered; transportation of the pallets from vehicle to store house, lifting the pallets to the shelf, resting of the stock on the shelf, getting down the pallets, and carrying the pallets to production area. Using MCNP-4C exposure situations have been simulated, considering that the source has a homogeneous composition, the minimum stock in the store house is constituted by 7 pallets, and the several distances between pallets and workers when they are at work. The photons flux obtained by MCNP-4C is multiplied by the conversion factor of Flux to Kerma for air by conversion factor to Effective Dose by Kerma unit, and by the number of emitted photons. Those conversion factors are obtained of ICRP 74 table 1 and table 17 respectively. This

  18. ZZ MCNPDATA, Standard Neutron, Photon and Electron Data Libraries for MCNP-4C and MCB1C

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description: These cross-section libraries are released by the Diagnostics Applications Group, X-5, at Los Alamos National Laboratory for use with the MCNP Monte Carlo code package. This release includes all of the X-5 distributed neutron data libraries, the photon libraries MCPLIB1 and MCPLIB02, the electron libraries EL1 and EL03, an updated XSDIR file, and information files Readme.txt and Readme e ndf60.txt. This release is intended to completely replace previous RSICC releases DLC-105, DLC-181, and DLC-189 as well as the cross sections previously included with CCC-200/MCNP4A, and will be updated as new libraries become available. The README file provides information regarding each data library of this release. Additional documentation for some of the individual libraries and example SPECS files for use with MAKXSF are also provided. The XSDIR file is specific to this release and may not work with previous packages. Currently the neutron data library ENDF60 (based on ENDF/B-VI, up through and including release 2) is the default library for continuous-energy neutron transport. Additionally, the libraries MCPLIB02 and EL03 are the default libraries for photon and electron transport respectively. More information on the data libraries contained in this release is available in Appendix G of the MCNP4C manual. 2 - Description of program or function: ZZ-MCB-DLC200 contains the same cross section tables as the DLC-0200/03 package for the MCNP-4C code, except that the installation procedures are adapted to the MCB1C code system (NEA 1643/01). 3 - Application of the data: DLC-200/MCNPDATA is for use with Version 4C and later of the MCNP transport code. This data library provides a comprehensive set of cross sections for a wide range of radiation transport applications using the Monte Carlo code package CCC-700/MCNP4C. See Appendix G of the MCNP report LA-13709-M for information on the libraries and how to select specific nuclides for use in MCNP. 4 - Source and scope

  19. Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

    International Nuclear Information System (INIS)

    Huda, M.Q.

    2006-01-01

    The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

  20. SU-E-T-212: Comparison of TG-43 Dosimetric Parameters of Low and High Energy Brachytherapy Sources Obtained by MCNP Code Versions of 4C, X and 5

    Energy Technology Data Exchange (ETDEWEB)

    Zehtabian, M; Zaker, N; Sina, S [Shiraz University, Shiraz, Fars (Iran, Islamic Republic of); Meigooni, A Soleimani [Comprehensive Cancer Center of Nevada, Las Vegas, Nevada (United States)

    2015-06-15

    Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 which is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.

  1. MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries

    International Nuclear Information System (INIS)

    Sublet, J.Ch.

    2006-01-01

    Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S(α,β) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)

  2. Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

    2004-07-01

    MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

  3. The MCNP-4C2 design of a two element photon/electron dosemeter that uses magnesium/copper/phosphorus doped lithium fluoride.

    Science.gov (United States)

    Eakins, J S; Bartlett, D T; Hager, L G; Molinos-Solsona, C; Tanner, R J

    2008-01-01

    The Health Protection Agency is changing from using detectors made from 7LiF:Mg,Ti in its photon/electron personal dosemeters, to 7LiF:Mg,Cu,P. Specifically, the Harshaw TLD-700H card is to be adopted. As a consequence of this change, the dosemeter holder is also being modified not only to accommodate the shape of the new card, but also to optimize the photon and electron response characteristics of the device. This redesign process was achieved using MCNP-4C2 and the kerma approximation, electron range/energy tables with additional electron transport calculations, and experimental validation, with different potential filters compared; the optimum filter studied was a polytetrafluoroethylene disc of diameter 18 mm and thickness 4.3 mm. Calculated relative response characteristics at different angles of incidence and energies between 16 and 6174 keV are presented for this new dosemeter configuration and compared with measured type-test results. A new estimate for the energy-dependent relative light conversion efficiency appropriate to the 7LiF:Mg,Cu,P was also derived for determining the correct dosemeter response.

  4. Comparison of the dose evaluation methods for criticality accident

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Oka, Tsutomu

    2004-01-01

    The improvement of the dose evaluation method for criticality accidents is important to rationalize design of the nuclear fuel cycle facilities. The source spectrums of neutron and gamma ray of a criticality accident depend on the condition of the source, its materials, moderation, density and so on. The comparison of the dose evaluation methods for a criticality accident is made. Some methods, which are combination of criticality calculation and shielding calculation, are proposed. Prompt neutron and gamma ray doses from nuclear criticality of some uranium systems have been evaluated as the Nuclear Criticality Slide Rule. The uranium metal source (unmoderated system) and the uranyl nitrate solution source (moderated system) in the rule are evaluated by some calculation methods, which are combinations of code and cross section library, as follows: (a) SAS1X (ENDF/B-IV), (b) MCNP4C (ENDF/B-VI)-ANISN (DLC23E or JSD120), (c) MCNP4C-MCNP4C (ENDF/B-VI). They have consisted of criticality calculation and shielding calculation. These calculation methods are compared about the tissue absorbed dose and the spectrums at 2 m from the source. (author)

  5. MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries; Les bibliotheques Eccolib-Jeff-3.1

    Energy Technology Data Exchange (ETDEWEB)

    Sublet, J.Ch

    2006-07-01

    Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S({alpha},{beta}) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)

  6. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    International Nuclear Information System (INIS)

    Pauzi, A M

    2013-01-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U -battery TM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  8. Progress in nuclear well logging modeling using deterministic transport codes

    International Nuclear Information System (INIS)

    Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.

    2002-01-01

    Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)

  9. Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabadi, M.N., E-mail: mnnasrabadi@ast.ui.ac.ir [Department of Nuclear Engineering, Faculty of Advanced Sciences and Technologies, University of Isfahan, Isfahan 81746-73441 (Iran, Islamic Republic of); Bakhshi, F.; Jalali, M.; Mohammadi, A. [Department of Nuclear Engineering, Faculty of Advanced Sciences and Technologies, University of Isfahan, Isfahan 81746-73441 (Iran, Islamic Republic of)

    2011-12-11

    Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma 10.8 MeV following radioactive neutron capture by {sup 14}N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A {sup 252}Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C{sub 3}H{sub 6}N{sub 6}). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.

  10. Automatic coding method of the ACR Code

    International Nuclear Information System (INIS)

    Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi

    1993-01-01

    The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology

  11. Design of tallying function for general purpose Monte Carlo particle transport code JMCT

    International Nuclear Information System (INIS)

    Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin

    2013-01-01

    A new postponed accumulation algorithm was proposed. Based on JCOGIN (J combinatorial geometry Monte Carlo transport infrastructure) framework and the postponed accumulation algorithm, the tallying function of the general purpose Monte Carlo neutron-photon transport code JMCT was improved markedly. JMCT gets a higher tallying efficiency than MCNP 4C by 28% for simple geometry model, and JMCT is faster than MCNP 4C by two orders of magnitude for complicated repeated structure model. The available ability of tallying function for JMCT makes firm foundation for reactor analysis and multi-step burnup calculation. (authors)

  12. Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2007-01-01

    High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

  13. The spammed code offset method

    NARCIS (Netherlands)

    Skoric, B.; Vreede, de N.

    2013-01-01

    Helper data schemes are a security primitive used for privacy-preserving biometric databases and Physical Unclonable Functions. One of the oldest known helper data schemes is the Code Offset Method (COM). We propose an extension of the COM: the helper data is accompanied by many instances of fake

  14. The spammed code offset method

    NARCIS (Netherlands)

    Skoric, B.; Vreede, de N.

    2014-01-01

    Helper data schemes are a security primitive used for privacy-preserving biometric databases and physical unclonable functions. One of the oldest known helper data schemes is the code offset method (COM). We propose an extension of the COM: the helper data are accompanied by many instances of fake

  15. An assessment of the MCNP4C weight window

    International Nuclear Information System (INIS)

    Culbertson, Christopher N.; Hendricks, John S.

    1999-01-01

    A new, enhanced weight window generator suite has been developed for MCNP version 4C. The new generator correctly estimates importances in either a user-specified, geometry-independent, orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. The new generator is applied in a set of five variance reduction problems. The improved generator is compared with the weight window generator applied in MCNP4B. The benefits of the new methodology are highlighted, along with a description of its limitations. The authors also provide recommendations for utilization of the weight window generator

  16. Application of Monte Carlo method for dose calculation in thyroid follicle

    International Nuclear Information System (INIS)

    Silva, Frank Sinatra Gomes da

    2008-02-01

    The Monte Carlo method is an important tool to simulate radioactive particles interaction with biologic medium. The principal advantage of the method when compared with deterministic methods is the ability to simulate a complex geometry. Several computational codes use the Monte Carlo method to simulate the particles transport and they have the capacity to simulate energy deposition in models of organs and/or tissues, as well models of cells of human body. Thus, the calculation of the absorbed dose to thyroid's follicles (compound of colloid and follicles' cells) have a fundamental importance to dosimetry, because these cells are radiosensitive due to ionizing radiation exposition, in particular, exposition due to radioisotopes of iodine, because a great amount of radioiodine may be released into the environment in case of a nuclear accidents. In this case, the goal of this work was use the code of particles transport MNCP4C to calculate absorbed doses in models of thyroid's follicles, for Auger electrons, internal conversion electrons and beta particles, by iodine-131 and short-lived iodines (131, 132, 133, 134 e 135), with diameters varying from 30 to 500 μm. The results obtained from simulation with the MCNP4C code shown an average percentage of the 25% of total absorbed dose by colloid to iodine- 131 and 75% to short-lived iodine's. For follicular cells, this percentage was of 13% to iodine-131 and 87% to short-lived iodine's. The contributions from particles with low energies, like Auger and internal conversion electrons should not be neglected, to assessment the absorbed dose in cellular level. Agglomerative hierarchical clustering was used to compare doses obtained by codes MCNP4C, EPOTRAN, EGS4 and by deterministic methods. (author)

  17. Monte Carlo method for calculating the radiation skyshine produced by electron accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Kong Chaocheng [Department of Engineering Physics, Tsinghua University Beijing 100084 (China)]. E-mail: kongchaocheng@tsinghua.org.cn; Li Quanfeng [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Chen Huaibi [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Du Taibin [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Cheng Cheng [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Tang Chuanxiang [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Zhu Li [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China); Zhang Hui [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China); Pei Zhigang [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China); Ming Shenjin [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China)

    2005-06-01

    Using the MCNP4C Monte Carlo code, the X-ray skyshine produced by 9 MeV, 15 MeV and 21 MeV electron linear accelerators were calculated respectively with a new two-step method combined with the split and roulette variance reduction technique. Results of the Monte Carlo simulation, the empirical formulas used for skyshine calculation and the dose measurements were analyzed and compared. In conclusion, the skyshine dose measurements agreed reasonably with the results computed by the Monte Carlo method, but deviated from computational results given by empirical formulas. The effect on skyshine dose caused by different structures of accelerator head is also discussed in this paper.

  18. New Channel Coding Methods for Satellite Communication

    Directory of Open Access Journals (Sweden)

    J. Sebesta

    2010-04-01

    Full Text Available This paper deals with the new progressive channel coding methods for short message transmission via satellite transponder using predetermined length of frame. The key benefits of this contribution are modification and implementation of a new turbo code and utilization of unique features with applications of methods for bit error rate estimation and algorithm for output message reconstruction. The mentioned methods allow an error free communication with very low Eb/N0 ratio and they have been adopted for satellite communication, however they can be applied for other systems working with very low Eb/N0 ratio.

  19. Modeling the irradiation facility in the Deir Al-Hajar area to calculate the spatial gamma dose distribution using the MCNP code

    International Nuclear Information System (INIS)

    Khattab, K.; Bush, M; Kassery, H.

    2009-03-01

    A 3-D model for the irradiation plant which belongs to the Atomic Energy Commission, Department of Radiation Technology in the Deir Al-Hajar area near Damascus, is presented in this work using the MCNP-4C code. This model is used to calculate the spatial gamma ray dose in the (x, y, z) coordinate. Good agreements are noticed between the measured and the calculated results. (author)

  20. Method for coding low entrophy data

    Science.gov (United States)

    Yeh, Pen-Shu (Inventor)

    1995-01-01

    A method of lossless data compression for efficient coding of an electronic signal of information sources of very low information rate is disclosed. In this method, S represents a non-negative source symbol set, (s(sub 0), s(sub 1), s(sub 2), ..., s(sub N-1)) of N symbols with s(sub i) = i. The difference between binary digital data is mapped into symbol set S. Consecutive symbols in symbol set S are then paired into a new symbol set Gamma which defines a non-negative symbol set containing the symbols (gamma(sub m)) obtained as the extension of the original symbol set S. These pairs are then mapped into a comma code which is defined as a coding scheme in which every codeword is terminated with the same comma pattern, such as a 1. This allows a direct coding and decoding of the n-bit positive integer digital data differences without the use of codebooks.

  1. Numerical method improvement for a subchannel code

    Energy Technology Data Exchange (ETDEWEB)

    Ding, W.J.; Gou, J.L.; Shan, J.Q. [Xi' an Jiaotong Univ., Shaanxi (China). School of Nuclear Science and Technology

    2016-07-15

    Previous studies showed that the subchannel codes need most CPU time to solve the matrix formed by the conservation equations. Traditional matrix solving method such as Gaussian elimination method and Gaussian-Seidel iteration method cannot meet the requirement of the computational efficiency. Therefore, a new algorithm for solving the block penta-diagonal matrix is designed based on Stone's incomplete LU (ILU) decomposition method. In the new algorithm, the original block penta-diagonal matrix will be decomposed into a block upper triangular matrix and a lower block triangular matrix as well as a nonzero small matrix. After that, the LU algorithm is applied to solve the matrix until the convergence. In order to compare the computational efficiency, the new designed algorithm is applied to the ATHAS code in this paper. The calculation results show that more than 80 % of the total CPU time can be saved with the new designed ILU algorithm for a 324-channel PWR assembly problem, compared with the original ATHAS code.

  2. A mathematical method to calculate efficiency of BF3 detectors

    International Nuclear Information System (INIS)

    Si Fenni; Hu Qingyuan; Peng Taiping

    2009-01-01

    In order to calculate absolute efficiency of the BF 3 detector, MCNP/4C code is applied to calculate relative efficiency of the BF 3 detector first, and then absolute efficiency is figured out through mathematical techniques. Finally an energy response curve of the BF 3 detector for 1-20 MeV neutrons is derived. It turns out that efficiency of BF 3 detector are relatively uniform for 2-16 MeV neutrons. (authors)

  3. Subband Coding Methods for Seismic Data Compression

    Science.gov (United States)

    Kiely, A.; Pollara, F.

    1995-01-01

    This paper presents a study of seismic data compression techniques and a compression algorithm based on subband coding. The compression technique described could be used as a progressive transmission system, where successive refinements of the data can be requested by the user. This allows seismologists to first examine a coarse version of waveforms with minimal usage of the channel and then decide where refinements are required. Rate-distortion performance results are presented and comparisons are made with two block transform methods.

  4. Monte Carlo determination of the conversion coefficients Hp(3)/Ka in a right cylinder phantom with 'PENELOPE' code. Comparison with 'MCNP' simulations.

    Science.gov (United States)

    Daures, J; Gouriou, J; Bordy, J M

    2011-03-01

    This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

  5. FAST PALMPRINT AUTHENTICATION BY SOBEL CODE METHOD

    Directory of Open Access Journals (Sweden)

    Jyoti Malik

    2011-05-01

    Full Text Available The ideal real time personal authentication system should be fast and accurate to automatically identify a person’s identity. In this paper, we have proposed a palmprint based biometric authentication method with improvement in time and accuracy, so as to make it a real time palmprint authentication system. Several edge detection methods, wavelet transform, phase congruency etc. are available to extract line feature from the palmprint. In this paper, Multi-scale Sobel Code operators of different orientations (0?, 45?, 90?, and 135? are applied to the palmprint to extract Sobel-Palmprint features in different direc- tions. The Sobel-Palmprint features extracted are stored in Sobel- Palmprint feature vector and matched using sliding window with Hamming Distance similarity measurement method. The sliding win- dow method is accurate but time taking process. In this paper, we have improved the sliding window method so that the matching time reduces. It is observed that there is 39.36% improvement in matching time. In addition, a Min Max Threshold Range (MMTR method is proposed that helps in increasing overall system accuracy by reducing the False Acceptance Rate (FAR. Experimental results indicate that the MMTR method improves the False Acceptance Rate drastically and improvement in sliding window method reduces the comparison time. The accuracy improvement and matching time improvement leads to proposed real time authentication system.

  6. Simulation of the BNCT of Brain Tumors Using MCNP Code: Beam Designing and Dose Evaluation

    Directory of Open Access Journals (Sweden)

    Fatemeh Sadat Rasouli

    2012-09-01

    Full Text Available Introduction BNCT is an effective method to destroy brain tumoral cells while sparing the healthy tissues. The recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. In this paper, it is indicated that using D-T neutron source and optimizing of Beam Shaping Assembly (BSA leads to treating brain tumors in a reasonable time where all IAEA recommended criteria are met. Materials and Methods The proposed BSA based on a D-T neutron generator consists of a neutron multiplier system, moderators, reflector, and collimator. The simulated Snyder head phantom is used to evaluate dose profiles in tissues due to the irradiation of designed beam. Monte Carlo Code, MCNP-4C, was used in order to perform these calculations.   Results The neutron beam associated with the designed and optimized BSA has an adequate epithermal flux at the beam port and neutron and gamma contaminations are removed as much as possible. Moreover, it was showed that increasing J/Φ, as a measure of beam directionality, leads to improvement of beam performance and survival of healthy tissues surrounding the tumor. Conclusion According to the simulation results, the proposed system based on D-T neutron source, which is suitable for in-hospital installations, satisfies all in-air parameters. Moreover, depth-dose curves investigate proper performance of designed beam in tissues. The results are comparable with the performances of other facilities.

  7. Evaluation of a special pencil ionization chamber by the Monte Carlo method

    International Nuclear Information System (INIS)

    Mendonca, Dalila; Neves, Lucio P.; Perini, Ana P.

    2015-01-01

    A special pencil type ionization chamber, developed at the Instituto de Pesquisas Energeticas e Nucleares, was characterized by means of Monte Carlo simulation to determine the influence of its components on its response. The main differences between this ionization chamber and commercial ionization chambers are related to its configuration and constituent materials. The simulations were made employing the MCNP-4C Monte Carlo code. The highest influence was obtained for the body of PMMA: 7.0%. (author)

  8. Use of the MCNP Monte Carlo code for characterization of a pencil-type ionization chamber; Uso do código de Monte Carlo MCNP para caracterização de uma câmara de ionização tipo lápis

    Energy Technology Data Exchange (ETDEWEB)

    Mendonça, Dalila Souza Costa; Santos, William S.; Perini, Ana Paula, E-mail: anapaula.perini@ufu.br [Universidade Federal de Uberlândia (INFIS/UFU), MG (Brazil). Instituto de Física; Neves, Lucio Pereira; Caldas, Linda V. E. [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Belinato, Walmir [Instituto Federal de Educação, Ciência e Tecnologia da Bahia (IFBA), Vitória da Conquista, BA (Brazil)

    2017-07-01

    Ionization chambers are widely used in diagnostic radiology dosimetry. In this work, a special pencil-type ionization chamber, with different dimensions, configuration and materials in relation to commercial ones, was studied computationally. For this, the MCNP-4C Monte Carlo code and different radiation spectra were used to determine the influence of its components on its response. It was possible to observe that the highest influence was for the PVC wall. (author)

  9. A Method for Improving the Progressive Image Coding Algorithms

    Directory of Open Access Journals (Sweden)

    Ovidiu COSMA

    2014-12-01

    Full Text Available This article presents a method for increasing the performance of the progressive coding algorithms for the subbands of images, by representing the coefficients with a code that reduces the truncation error.

  10. A novel method of generating and remembering international morse codes

    Digital Repository Service at National Institute of Oceanography (India)

    Charyulu, R.J.K.

    untethered communications have been advanced, despite as S.O.S International Morse Code will be at rescue as an emergency tool, when all other modes fail The details of hte method and actual codes have been enumerated....

  11. Comparison of TG‐43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes

    Science.gov (United States)

    Zaker, Neda; Sina, Sedigheh; Koontz, Craig; Meigooni1, Ali S.

    2016-01-01

    Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross‐sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross‐sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in  125I and  103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP codeMCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low‐energy sources such as  125I and  103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for  103Pd and 10 cm for  125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for  192Ir and less than 1.2% for  137Cs between the three codes. PACS number(s): 87.56.bg PMID:27074460

  12. Comparison of TG-43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes.

    Science.gov (United States)

    Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S

    2016-03-08

    Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes.

  13. A method for scientific code coupling in a distributed environment

    International Nuclear Information System (INIS)

    Caremoli, C.; Beaucourt, D.; Chen, O.; Nicolas, G.; Peniguel, C.; Rascle, P.; Richard, N.; Thai Van, D.; Yessayan, A.

    1994-12-01

    This guide book deals with coupling of big scientific codes. First, the context is introduced: big scientific codes devoted to a specific discipline coming to maturity, and more and more needs in terms of multi discipline studies. Then we describe different kinds of code coupling and an example of code coupling: 3D thermal-hydraulic code THYC and 3D neutronics code COCCINELLE. With this example we identify problems to be solved to realize a coupling. We present the different numerical methods usable for the resolution of coupling terms. This leads to define two kinds of coupling: with the leak coupling, we can use explicit methods, and with the strong coupling we need to use implicit methods. On both cases, we analyze the link with the way of parallelizing code. For translation of data from one code to another, we define the notion of Standard Coupling Interface based on a general structure for data. This general structure constitutes an intermediary between the codes, thus allowing a relative independence of the codes from a specific coupling. The proposed method for the implementation of a coupling leads to a simultaneous run of the different codes, while they exchange data. Two kinds of data communication with message exchange are proposed: direct communication between codes with the use of PVM product (Parallel Virtual Machine) and indirect communication with a coupling tool. This second way, with a general code coupling tool, is based on a coupling method, and we strongly recommended to use it. This method is based on the two following principles: re-usability, that means few modifications on existing codes, and definition of a code usable for coupling, that leads to separate the design of a code usable for coupling from the realization of a specific coupling. This coupling tool available from beginning of 1994 is described in general terms. (authors). figs., tabs

  14. A comparison study for mass attenuation coefficients of some amino acids using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Vahabi, Seyed Milad; Bahreynipour, Mostean; Shamsaie-Zafarghandi, Mojtaba [Amirkabir Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this study, a novel model of MCNP4C code reported recently was used to determine the photon mass attenuation coefficients of some amino acids at energies, 123, 360, 511, 662, 1170, 1280 and 1330 keV. The simulation results were compared with the XCOM data. It was indicated that the results were highly close to the calculated XCOM values. Obtained results were used to calculate the molar extinction coefficient. All the results showed the convenience and usefulness of the model in calculation of mass attenuation coefficients of amino acids.

  15. EchoSeed Model 6733 Iodine-125 brachytherapy source: Improved dosimetric characterization using the MCNP5 Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Mosleh-Shirazi, M. A.; Hadad, K.; Faghihi, R.; Baradaran-Ghahfarokhi, M.; Naghshnezhad, Z.; Meigooni, A. S. [Center for Research in Medical Physics and Biomedical Engineering and Physics Unit, Radiotherapy Department, Shiraz University of Medical Sciences, Shiraz 71936-13311 (Iran, Islamic Republic of); Radiation Research Center and Medical Radiation Department, School of Engineering, Shiraz University, Shiraz 71936-13311 (Iran, Islamic Republic of); Comprehensive Cancer Center of Nevada, Las Vegas, Nevada 89169 (United States)

    2012-08-15

    This study primarily aimed to obtain the dosimetric characteristics of the Model 6733 {sup 125}I seed (EchoSeed) with improved precision and accuracy using a more up-to-date Monte-Carlo code and data (MCNP5) compared to previously published results, including an uncertainty analysis. Its secondary aim was to compare the results obtained using the MCNP5, MCNP4c2, and PTRAN codes for simulation of this low-energy photon-emitting source. The EchoSeed geometry and chemical compositions together with a published {sup 125}I spectrum were used to perform dosimetric characterization of this source as per the updated AAPM TG-43 protocol. These simulations were performed in liquid water material in order to obtain the clinically applicable dosimetric parameters for this source model. Dose rate constants in liquid water, derived from MCNP4c2 and MCNP5 simulations, were found to be 0.993 cGyh{sup -1} U{sup -1} ({+-}1.73%) and 0.965 cGyh{sup -1} U{sup -1} ({+-}1.68%), respectively. Overall, the MCNP5 derived radial dose and 2D anisotropy functions results were generally closer to the measured data (within {+-}4%) than MCNP4c and the published data for PTRAN code (Version 7.43), while the opposite was seen for dose rate constant. The generally improved MCNP5 Monte Carlo simulation may be attributed to a more recent and accurate cross-section library. However, some of the data points in the results obtained from the above-mentioned Monte Carlo codes showed no statistically significant differences. Derived dosimetric characteristics in liquid water are provided for clinical applications of this source model.

  16. Calibration Methods for Reliability-Based Design Codes

    DEFF Research Database (Denmark)

    Gayton, N.; Mohamed, A.; Sørensen, John Dalsgaard

    2004-01-01

    The calibration methods are applied to define the optimal code format according to some target safety levels. The calibration procedure can be seen as a specific optimization process where the control variables are the partial factors of the code. Different methods are available in the literature...

  17. Lattice Boltzmann method fundamentals and engineering applications with computer codes

    CERN Document Server

    Mohamad, A A

    2014-01-01

    Introducing the Lattice Boltzmann Method in a readable manner, this book provides detailed examples with complete computer codes. It avoids the most complicated mathematics and physics without scarifying the basic fundamentals of the method.

  18. Statistical methods for accurately determining criticality code bias

    International Nuclear Information System (INIS)

    Trumble, E.F.; Kimball, K.D.

    1997-01-01

    A system of statistically treating validation calculations for the purpose of determining computer code bias is provided in this paper. The following statistical treatments are described: weighted regression analysis, lower tolerance limit, lower tolerance band, and lower confidence band. These methods meet the criticality code validation requirements of ANS 8.1. 8 refs., 5 figs., 4 tabs

  19. Control rod computer code IAMCOS: general theory and numerical methods

    International Nuclear Information System (INIS)

    West, G.

    1982-11-01

    IAMCOS is a computer code for the description of mechanical and thermal behavior of cylindrical control rods for fast breeders. This code version was applied, tested and modified from 1979 to 1981. In this report are described the basic model (02 version), theoretical definitions and computation methods [fr

  20. Method and device for decoding coded digital video signals

    NARCIS (Netherlands)

    2000-01-01

    The invention relates to a video coding method and system including a quantization and coding sub-assembly (38) in which a quantization parameter is controlled by another parameter defined as being in direct relation with the dynamic range value of the data contained in given blocks of pixels.

  1. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  2. Advanced codes and methods supporting improved fuel cycle economics - 5493

    International Nuclear Information System (INIS)

    Curca-Tivig, F.; Maupin, K.; Thareau, S.

    2015-01-01

    AREVA's code development program was practically completed in 2014. The basic codes supporting a new generation of advanced methods are the followings. GALILEO is a state-of-the-art fuel rod performance code for PWR and BWR applications. Development is completed, implementation started in France and the U.S.A. ARCADIA-1 is a state-of-the-art neutronics/ thermal-hydraulics/ thermal-mechanics code system for PWR applications. Development is completed, implementation started in Europe and in the U.S.A. The system thermal-hydraulic codes S-RELAP5 and CATHARE-2 are not really new but still state-of-the-art in the domain. S-RELAP5 was completely restructured and re-coded such that its life cycle increases by further decades. CATHARE-2 will be replaced in the future by the new CATHARE-3. The new AREVA codes and methods are largely based on first principles modeling with an extremely broad international verification and validation data base. This enables AREVA and its customers to access more predictable licensing processes in a fast evolving regulatory environment (new safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation...). In this context, the advanced codes and methods and the associated verification and validation represent the key to avoiding penalties on products, on operational limits, or on methodologies themselves

  3. The variational celular method - the code implantation

    International Nuclear Information System (INIS)

    Rosato, A.; Lima, M.A.P.

    1980-12-01

    The process to determine the potential energy curve for diatomic molecules by the Variational Cellular Method is discussed. An analysis of the determination of the electronic eigenenergies and the electrostatic energy of these molecules is made. An explanation of the input data and their meaning is also presented. (Author) [pt

  4. New decoding methods of interleaved burst error-correcting codes

    Science.gov (United States)

    Nakano, Y.; Kasahara, M.; Namekawa, T.

    1983-04-01

    A probabilistic method of single burst error correction, using the syndrome correlation of subcodes which constitute the interleaved code, is presented. This method makes it possible to realize a high capability of burst error correction with less decoding delay. By generalizing this method it is possible to obtain probabilistic method of multiple (m-fold) burst error correction. After estimating the burst error positions using syndrome correlation of subcodes which are interleaved m-fold burst error detecting codes, this second method corrects erasure errors in each subcode and m-fold burst errors. The performance of these two methods is analyzed via computer simulation, and their effectiveness is demonstrated.

  5. Structural reliability methods: Code development status

    Science.gov (United States)

    Millwater, Harry R.; Thacker, Ben H.; Wu, Y.-T.; Cruse, T. A.

    1991-05-01

    The Probabilistic Structures Analysis Method (PSAM) program integrates state of the art probabilistic algorithms with structural analysis methods in order to quantify the behavior of Space Shuttle Main Engine structures subject to uncertain loadings, boundary conditions, material parameters, and geometric conditions. An advanced, efficient probabilistic structural analysis software program, NESSUS (Numerical Evaluation of Stochastic Structures Under Stress) was developed as a deliverable. NESSUS contains a number of integrated software components to perform probabilistic analysis of complex structures. A nonlinear finite element module NESSUS/FEM is used to model the structure and obtain structural sensitivities. Some of the capabilities of NESSUS/FEM are shown. A Fast Probability Integration module NESSUS/FPI estimates the probability given the structural sensitivities. A driver module, PFEM, couples the FEM and FPI. NESSUS, version 5.0, addresses component reliability, resistance, and risk.

  6. A code for obtaining temperature distribution by finite element method

    International Nuclear Information System (INIS)

    Bloch, M.

    1984-01-01

    The ELEFIB Fortran language computer code using finite element method for calculating temperature distribution of linear and two dimensional problems, in permanent region or in the transient phase of heat transfer, is presented. The formulation of equations uses the Galerkin method. Some examples are shown and the results are compared with other papers. The comparative evaluation shows that the elaborated code gives good values. (M.C.K.) [pt

  7. Development and application of methods to characterize code uncertainty

    International Nuclear Information System (INIS)

    Wilson, G.E.; Burtt, J.D.; Case, G.S.; Einerson, J.J.; Hanson, R.G.

    1985-01-01

    The United States Nuclear Regulatory Commission sponsors both international and domestic studies to assess its safety analysis codes. The Commission staff intends to use the results of these studies to quantify the uncertainty of the codes with a statistically based analysis method. Development of the methodology is underway. The Idaho National Engineering Laboratory contributions to the early development effort, and testing of two candidate methods are the subjects of this paper

  8. The use of the MCNP code for the quantitative analysis of elements in geological formations

    Energy Technology Data Exchange (ETDEWEB)

    Cywicka-Jakiel, T.; Woynicka, U. [The Henryk Niewodniczanski Institute of Nuclear Physics, Krakow (Poland); Zorski, T. [University of Mining and Metallurgy, Faculty of Geology, Geophysics and Environmental Protection, Krakow (Poland)

    2003-07-01

    The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)

  9. The use of the MCNP code for the quantitative analysis of elements in geological formations

    International Nuclear Information System (INIS)

    Cywicka-Jakiel, T.; Woynicka, U.; Zorski, T.

    2003-01-01

    The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)

  10. Methods and computer codes for probabilistic sensitivity and uncertainty analysis

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1985-01-01

    This paper describes the methods and applications experience with two computer codes that are now available from the National Energy Software Center at Argonne National Laboratory. The purpose of the SCREEN code is to identify a group of most important input variables of a code that has many (tens, hundreds) input variables with uncertainties, and do this without relying on judgment or exhaustive sensitivity studies. Purpose of the PROSA-2 code is to propagate uncertainties and calculate the distributions of interesting output variable(s) of a safety analysis code using response surface techniques, based on the same runs used for screening. Several applications are discussed, but the codes are generic, not tailored to any specific safety application code. They are compatible in terms of input/output requirements but also independent of each other, e.g., PROSA-2 can be used without first using SCREEN if a set of important input variables has first been selected by other methods. Also, although SCREEN can select cases to be run (by random sampling), a user can select cases by other methods if he so prefers, and still use the rest of SCREEN for identifying important input variables

  11. Parallelization methods study of thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Gaudart, Catherine

    2000-01-01

    The variety of parallelization methods and machines leads to a wide selection for programmers. In this study we suggest, in an industrial context, some solutions from the experience acquired through different parallelization methods. The study is about several scientific codes which simulate a large variety of thermal-hydraulics phenomena. A bibliography on parallelization methods and a first analysis of the codes showed the difficulty of our process on the whole applications to study. Therefore, it would be necessary to identify and extract a representative part of these applications and parallelization methods. The linear solver part of the codes forced itself. On this particular part several parallelization methods had been used. From these developments one could estimate the necessary work for a non initiate programmer to parallelize his application, and the impact of the development constraints. The different methods of parallelization tested are the numerical library PETSc, the parallelizer PAF, the language HPF, the formalism PEI and the communications library MPI and PYM. In order to test several methods on different applications and to follow the constraint of minimization of the modifications in codes, a tool called SPS (Server of Parallel Solvers) had be developed. We propose to describe the different constraints about the optimization of codes in an industrial context, to present the solutions given by the tool SPS, to show the development of the linear solver part with the tested parallelization methods and lastly to compare the results against the imposed criteria. (author) [fr

  12. A GPU code for analytic continuation through a sampling method

    Directory of Open Access Journals (Sweden)

    Johan Nordström

    2016-01-01

    Full Text Available We here present a code for performing analytic continuation of fermionic Green’s functions and self-energies as well as bosonic susceptibilities on a graphics processing unit (GPU. The code is based on the sampling method introduced by Mishchenko et al. (2000, and is written for the widely used CUDA platform from NVidia. Detailed scaling tests are presented, for two different GPUs, in order to highlight the advantages of this code with respect to standard CPU computations. Finally, as an example of possible applications, we provide the analytic continuation of model Gaussian functions, as well as more realistic test cases from many-body physics.

  13. 2D arc-PIC code description: methods and documentation

    CERN Document Server

    Timko, Helga

    2011-01-01

    Vacuum discharges are one of the main limiting factors for future linear collider designs such as that of the Compact LInear Collider. To optimize machine efficiency, maintaining the highest feasible accelerating gradient below a certain breakdown rate is desirable; understanding breakdowns can therefore help us to achieve this goal. As a part of ongoing theoretical research on vacuum discharges at the Helsinki Institute of Physics, the build-up of plasma can be investigated through the particle-in-cell method. For this purpose, we have developed the 2D Arc-PIC code introduced here. We present an exhaustive description of the 2D Arc-PIC code in two parts. In the first part, we introduce the particle-in-cell method in general and detail the techniques used in the code. In the second part, we provide a documentation and derivation of the key equations occurring in the code. The code is original work of the author, written in 2010, and is therefore under the copyright of the author. The development of the code h...

  14. A Fast Optimization Method for General Binary Code Learning.

    Science.gov (United States)

    Shen, Fumin; Zhou, Xiang; Yang, Yang; Song, Jingkuan; Shen, Heng; Tao, Dacheng

    2016-09-22

    Hashing or binary code learning has been recognized to accomplish efficient near neighbor search, and has thus attracted broad interests in recent retrieval, vision and learning studies. One main challenge of learning to hash arises from the involvement of discrete variables in binary code optimization. While the widely-used continuous relaxation may achieve high learning efficiency, the pursued codes are typically less effective due to accumulated quantization error. In this work, we propose a novel binary code optimization method, dubbed Discrete Proximal Linearized Minimization (DPLM), which directly handles the discrete constraints during the learning process. Specifically, the discrete (thus nonsmooth nonconvex) problem is reformulated as minimizing the sum of a smooth loss term with a nonsmooth indicator function. The obtained problem is then efficiently solved by an iterative procedure with each iteration admitting an analytical discrete solution, which is thus shown to converge very fast. In addition, the proposed method supports a large family of empirical loss functions, which is particularly instantiated in this work by both a supervised and an unsupervised hashing losses, together with the bits uncorrelation and balance constraints. In particular, the proposed DPLM with a supervised `2 loss encodes the whole NUS-WIDE database into 64-bit binary codes within 10 seconds on a standard desktop computer. The proposed approach is extensively evaluated on several large-scale datasets and the generated binary codes are shown to achieve very promising results on both retrieval and classification tasks.

  15. Dose mapping simulation using the MCNP code for the Syrian gamma irradiation facility and benchmarking

    International Nuclear Information System (INIS)

    Khattab, K.; Boush, M.; Alkassiri, H.

    2013-01-01

    Highlights: • The MCNP4C was used to calculate the gamma ray dose rate spatial distribution in for the SGIF. • Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter was conducted as well. • Good agreements were noticed between the calculated and measured results. • The maximum relative differences were less than 7%, 4% and 4% in the x, y and z directions respectively. - Abstract: A three dimensional model for the Syrian gamma irradiation facility (SGIF) is developed in this paper to calculate the gamma ray dose rate spatial distribution in the irradiation room at the 60 Co source board using the MCNP-4C code. Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter is conducted as well to compare the calculated and measured results. Good agreements are noticed between the calculated and measured results with maximum relative differences less than 7%, 4% and 4% in the x, y and z directions respectively. This agreement indicates that the established model is an accurate representation of the SGIF and can be used in the future to make the calculation design for a new irradiation facility

  16. Calculation of the X-Ray Spectrum of a Mammography System with Various Voltages and Different Anode-Filter Combinations Using MCNP Code

    Directory of Open Access Journals (Sweden)

    Lida Gholamkar

    2016-09-01

    Full Text Available Introduction One of the best methods in the diagnosis and control of breast cancer is mammography. The importance of mammography is directly related to its value in the detection of breast cancer in the early stages, which leads to a more effective treatment. The purpose of this article was to calculate the X-ray spectrum in a mammography system with Monte Carlo codes, including MCNPX and MCNP5. Materials and Methods The device, simulated using the MCNP code, was Planmed Nuance digital mammography device (Planmed Oy, Finland, equipped with an amorphous selenium detector. Different anode/filter materials, such as molybdenum-rhodium (Mo-Rh, molybdenum-molybdenum (Mo-Mo, tungsten-tin (W-Sn, tungsten-silver (W-Ag, tungsten-palladium (W-Pd, tungsten-aluminum (W-Al, tungsten-molybdenum (W-Mo, molybdenum-aluminum (Mo-Al, tungsten-rhodium (W-Rh, rhodium-aluminum (Rh-Al, and rhodium-rhodium (Rh-Rh, were simulated in this study. The voltage range of the X-ray tube was between 24 and 34 kV with a 2 kV interval. Results The charts of changing photon flux versus energy were plotted for different types of anode-filter combinations. The comparison with the findings reported by others indicated acceptable consistency. Also, the X-ray spectra, obtained from MCNP5 and MCNPX codes for W-Ag and W-Rh combinations, were compared. We compared the present results with the reported data of MCNP4C and IPEM report No. 78 for Mo-Mo, Mo-Rh, and W-Al combinations. Conclusion The MCNPX calculation outcomes showed acceptable results in a low-energy X-ray beam range (10-35 keV. The obtained simulated spectra for different anode/filter combinations were in good conformity with the finding of previous research.

  17. Combined backscatter and transmission method for nuclear density gauge

    Directory of Open Access Journals (Sweden)

    Golgoun Seyed Mohammad

    2015-01-01

    Full Text Available Nowadays, the use of nuclear density gauges, due to the ability to work in harsh industrial environments, is very common. In this study, to reduce error related to the ρ of continuous measuring density, the combination of backscatter and transmission are used simultaneously. For this reason, a 137Cs source for Compton scattering dominance and two detectors are simulated by MCNP4C code for measuring the density of 3 materials. Important advantages of this combined radiometric gauge are diminished influence of μ and therefore improving linear regression.

  18. Beam neutron energy optimization for boron neutron capture therapy using monte Carlo method

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Shekarian, E.

    2006-01-01

    In last two decades the optimal neutron energy for the treatment of deep seated tumors in boron neutron capture therapy in view of neutron physics and chemical compounds of boron carrier has been under thorough study. Although neutron absorption cross section of boron is high (3836b), the treatment of deep seated tumors such as glioblastoma multiform requires beam of neutrons of higher energy that can penetrate deeply into the brain and thermalized in the proximity of the tumor. Dosage from recoil proton associated with fast neutrons however poses some constraints on maximum neutron energy that can be used in the treatment. For this reason neutrons in the epithermal energy range of 10eV-10keV are generally to be the most appropriate. The simulation carried out by Monte Carlo methods using MCBNCT and MCNP4C codes along with the cross section library in 290 groups extracted from ENDF/B6 main library. The ptimal neutron energy for deep seated tumors depends on the sue and depth of tumor. Our estimated optimized energy for the tumor of 5cm wide and 1-2cm thick stands at 5cm depth is in the range of 3-5keV

  19. Automated uncertainty analysis methods in the FRAP computer codes

    International Nuclear Information System (INIS)

    Peck, S.O.

    1980-01-01

    A user oriented, automated uncertainty analysis capability has been incorporated in the Fuel Rod Analysis Program (FRAP) computer codes. The FRAP codes have been developed for the analysis of Light Water Reactor fuel rod behavior during steady state (FRAPCON) and transient (FRAP-T) conditions as part of the United States Nuclear Regulatory Commission's Water Reactor Safety Research Program. The objective of uncertainty analysis of these codes is to obtain estimates of the uncertainty in computed outputs of the codes is to obtain estimates of the uncertainty in computed outputs of the codes as a function of known uncertainties in input variables. This paper presents the methods used to generate an uncertainty analysis of a large computer code, discusses the assumptions that are made, and shows techniques for testing them. An uncertainty analysis of FRAP-T calculated fuel rod behavior during a hypothetical loss-of-coolant transient is presented as an example and carried through the discussion to illustrate the various concepts

  20. Deep Learning Methods for Improved Decoding of Linear Codes

    Science.gov (United States)

    Nachmani, Eliya; Marciano, Elad; Lugosch, Loren; Gross, Warren J.; Burshtein, David; Be'ery, Yair

    2018-02-01

    The problem of low complexity, close to optimal, channel decoding of linear codes with short to moderate block length is considered. It is shown that deep learning methods can be used to improve a standard belief propagation decoder, despite the large example space. Similar improvements are obtained for the min-sum algorithm. It is also shown that tying the parameters of the decoders across iterations, so as to form a recurrent neural network architecture, can be implemented with comparable results. The advantage is that significantly less parameters are required. We also introduce a recurrent neural decoder architecture based on the method of successive relaxation. Improvements over standard belief propagation are also observed on sparser Tanner graph representations of the codes. Furthermore, we demonstrate that the neural belief propagation decoder can be used to improve the performance, or alternatively reduce the computational complexity, of a close to optimal decoder of short BCH codes.

  1. MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted

  2. Optimization of the beam shaping assembly in the D-D neutron generators-based BNCT using the response matrix method.

    Science.gov (United States)

    Kasesaz, Y; Khalafi, H; Rahmani, F

    2013-12-01

    Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45 MeV neutrons that are produced in the D-D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70 cm Fluental as a moderator, 30 cm Pb as a reflector, 2mm (6)Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method. Copyright © 2013 Elsevier Ltd. All rights reserved.

  3. Research on coding and decoding method for digital levels

    Energy Technology Data Exchange (ETDEWEB)

    Tu Lifen; Zhong Sidong

    2011-01-20

    A new coding and decoding method for digital levels is proposed. It is based on an area-array CCD sensor and adopts mixed coding technology. By taking advantage of redundant information in a digital image signal, the contradiction that the field of view and image resolution restrict each other in a digital level measurement is overcome, and the geodetic leveling becomes easier. The experimental results demonstrate that the uncertainty of measurement is 1mm when the measuring range is between 2m and 100m, which can meet practical needs.

  4. Research on coding and decoding method for digital levels.

    Science.gov (United States)

    Tu, Li-fen; Zhong, Si-dong

    2011-01-20

    A new coding and decoding method for digital levels is proposed. It is based on an area-array CCD sensor and adopts mixed coding technology. By taking advantage of redundant information in a digital image signal, the contradiction that the field of view and image resolution restrict each other in a digital level measurement is overcome, and the geodetic leveling becomes easier. The experimental results demonstrate that the uncertainty of measurement is 1 mm when the measuring range is between 2 m and 100 m, which can meet practical needs.

  5. Optimized iterative decoding method for TPC coded CPM

    Science.gov (United States)

    Ma, Yanmin; Lai, Penghui; Wang, Shilian; Xie, Shunqin; Zhang, Wei

    2018-05-01

    Turbo Product Code (TPC) coded Continuous Phase Modulation (CPM) system (TPC-CPM) has been widely used in aeronautical telemetry and satellite communication. This paper mainly investigates the improvement and optimization on the TPC-CPM system. We first add the interleaver and deinterleaver to the TPC-CPM system, and then establish an iterative system to iteratively decode. However, the improved system has a poor convergence ability. To overcome this issue, we use the Extrinsic Information Transfer (EXIT) analysis to find the optimal factors for the system. The experiments show our method is efficient to improve the convergence performance.

  6. Application of Monte Carlo method in study of the padronization for radionuclides with complex disintegration scheme in 4πβ-γ coincidence System

    International Nuclear Information System (INIS)

    Takeda, Mauro Noriaki

    2006-01-01

    The present work described a new methodology for modelling the behaviour of the activity in a 4πβ-γ coincidence system. The detection efficiency for electrons in the proportional counter and gamma radiation in the NaI(Tl) detector was calculated using the Monte Carlo program MCNP4C. Another Monte Carlo code was developed which follows the path in the disintegration scheme from the initial state of the precursor radionuclide, until the ground state of the daughter nucleus. Every step of the disintegration scheme is sorted by random numbers taking into account the probabilities of all β - branches, electronic capture branches, transitions probabilities and internal conversion coefficients. Once the final state was reached beta, electronic capture events and gamma transitions are accounted for the three spectra: beta, gamma and coincidence variation in the beta efficiency was performed simulating energy cut off or use of absorbers (Collodion). The selected radionuclides for simulation were: 134 Cs, 72 Ga which disintegrate by β - transition, 133 Ba which disintegrates by electronic capture and 35 S which is a beta pure emitter. For the latter, the Efficiency Tracing technique was simulated. The extrapolation curves obtained by Monte Carlo were filled by the Least Square Method with the experimental points and the results were compared to the Linear Extrapolation method. (author)

  7. Method of laser beam coding for control systems

    Science.gov (United States)

    Pałys, Tomasz; Arciuch, Artur; Walczak, Andrzej; Murawski, Krzysztof

    2017-08-01

    The article presents the method of encoding a laser beam for control systems. The experiments were performed using a red laser emitting source with a wavelength of λ = 650 nm and a power of P ≍ 3 mW. The aim of the study was to develop methods of modulation and demodulation of the laser beam. Results of research, in which we determined the effect of selected camera parameters, such as image resolution, number of frames per second on the result of demodulation of optical signal, is also shown in the paper. The experiments showed that the adopted coding method provides sufficient information encoded in a single laser beam (36 codes with the effectiveness of decoding at 99.9%).

  8. Monte Carlo burnup codes acceleration using the correlated sampling method

    International Nuclear Information System (INIS)

    Dieudonne, C.

    2013-01-01

    For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step. In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4 code will be discussed, as well as the precise calculation scheme used to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes. (author) [fr

  9. CNC LATHE MACHINE PRODUCING NC CODE BY USING DIALOG METHOD

    Directory of Open Access Journals (Sweden)

    Yakup TURGUT

    2004-03-01

    Full Text Available In this study, an NC code generation program utilising Dialog Method was developed for turning centres. Initially, CNC lathes turning methods and tool path development techniques were reviewed briefly. By using geometric definition methods, tool path was generated and CNC part program was developed for FANUC control unit. The developed program made CNC part program generation process easy. The program was developed using BASIC 6.0 programming language while the material and cutting tool database were and supported with the help of ACCESS 7.0.

  10. New computational methods used in the lattice code DRAGON

    International Nuclear Information System (INIS)

    Marleau, G.; Hebert, A.; Roy, R.

    1992-01-01

    The lattice code DRAGON is used to perform transport calculations inside cells and assemblies for multidimensional geometry using the collision probability method, including the interface current and J ± techniques. Typical geometries that can be treated using this code include CANDU 2-dimensional clusters, CANDU 3-dimensional assemblies, pressurized water reactor (PWR) rectangular and hexagonal assemblies. It contains a self-shielding module for the treatment of microscopic cross section libraries and a depletion module for burnup calculations. DRAGON was written in a modular form in such a way as to accept easily new collision probability options and make them readily available to all the modules that require collision probability matrices like the self-shielding module, the flux solution module and the homogenization module. In this paper the authors present an overview of DRAGON and discuss some of the methods that were implemented in DRAGON in order to improve on its performance

  11. Computer codes and methods for simulating accelerator driven systems

    International Nuclear Information System (INIS)

    Sartori, E.; Byung Chan Na

    2003-01-01

    A large set of computer codes and associated data libraries have been developed by nuclear research and industry over the past half century. A large number of them are in the public domain and can be obtained under agreed conditions from different Information Centres. The areas covered comprise: basic nuclear data and models, reactor spectra and cell calculations, static and dynamic reactor analysis, criticality, radiation shielding, dosimetry and material damage, fuel behaviour, safety and hazard analysis, heat conduction and fluid flow in reactor systems, spent fuel and waste management (handling, transportation, and storage), economics of fuel cycles, impact on the environment of nuclear activities etc. These codes and models have been developed mostly for critical systems used for research or power generation and other technological applications. Many of them have not been designed for accelerator driven systems (ADS), but with competent use, they can be used for studying such systems or can form the basis for adapting existing methods to the specific needs of ADS's. The present paper describes the types of methods, codes and associated data available and their role in the applications. It provides Web addresses for facilitating searches for such tools. Some indications are given on the effect of non appropriate or 'blind' use of existing tools to ADS. Reference is made to available experimental data that can be used for validating the methods use. Finally, some international activities linked to the different computational aspects are described briefly. (author)

  12. Core physics design calculation of mini-type fast reactor based on Monte Carlo method

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2007-01-01

    An accurate physics calculation model has been set up for the mini-type sodium-cooled fast reactor (MFR) based on MCNP-4C code, then a detailed calculation of its critical physics characteristics, neutron flux distribution, power distribution and reactivity control has been carried out. The results indicate that the basic physics characteristics of MFR can satisfy the requirement and objectives of the core design. The power density and neutron flux distribution are symmetrical and reasonable. The control system is able to make a reliable reactivity balance efficiently and meets the request for long-playing operation. (authors)

  13. Vectorization, parallelization and porting of nuclear codes. 2001

    International Nuclear Information System (INIS)

    Akiyama, Mitsunaga; Katakura, Fumishige; Kume, Etsuo; Nemoto, Toshiyuki; Tsuruoka, Takuya; Adachi, Masaaki

    2003-07-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the super computer system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 10 codes in fiscal 2001. In this report, the parallelization of Neutron Radiography for 3 Dimensional CT code NR3DCT, the vectorization of unsteady-state heat conduction code THERMO3D, the porting of initial program of MHD simulation, the tuning of Heat And Mass Balance Analysis Code HAMBAC, the porting and parallelization of Monte Carlo N-Particle transport code MCNP4C3, the porting and parallelization of Monte Carlo N-Particle transport code system MCNPX2.1.5, the porting of induced activity calculation code CINAC-V4, the use of VisLink library in multidimensional two-fluid model code ACD3D and the porting of experiment data processing code from GS8500 to SR8000 are described. (author)

  14. CFD code verification and the method of manufactured solutions

    International Nuclear Information System (INIS)

    Pelletier, D.; Roache, P.J.

    2002-01-01

    This paper presents the Method of Manufactured Solutions (MMS) for CFD code verification. The MMS provides benchmark solutions for direct evaluation of the solution error. The best benchmarks are exact analytical solutions with sufficiently complex solution structure to ensure that all terms of the differential equations are exercised in the simulation. The MMS provides a straight forward and general procedure for generating such solutions. When used with systematic grid refinement studies, which are remarkably sensitive, the MMS provides strong code verification with a theorem-like quality. The MMS is first presented on simple 1-D examples. Manufactured solutions for more complex problems are then presented with sample results from grid convergence studies. (author)

  15. The impact of three discharge coding methods on the accuracy of diagnostic coding and hospital reimbursement for inpatient medical care.

    Science.gov (United States)

    Tsopra, Rosy; Peckham, Daniel; Beirne, Paul; Rodger, Kirsty; Callister, Matthew; White, Helen; Jais, Jean-Philippe; Ghosh, Dipansu; Whitaker, Paul; Clifton, Ian J; Wyatt, Jeremy C

    2018-07-01

    Coding of diagnoses is important for patient care, hospital management and research. However coding accuracy is often poor and may reflect methods of coding. This study investigates the impact of three alternative coding methods on the inaccuracy of diagnosis codes and hospital reimbursement. Comparisons of coding inaccuracy were made between a list of coded diagnoses obtained by a coder using (i)the discharge summary alone, (ii)case notes and discharge summary, and (iii)discharge summary with the addition of medical input. For each method, inaccuracy was determined for the primary, secondary diagnoses, Healthcare Resource Group (HRG) and estimated hospital reimbursement. These data were then compared with a gold standard derived by a consultant and coder. 107 consecutive patient discharges were analysed. Inaccuracy of diagnosis codes was highest when a coder used the discharge summary alone, and decreased significantly when the coder used the case notes (70% vs 58% respectively, p coded from the discharge summary with medical support (70% vs 60% respectively, p coding with case notes, and 35% for coding with medical support. The three coding methods resulted in an annual estimated loss of hospital remuneration of between £1.8 M and £16.5 M. The accuracy of diagnosis codes and percentage of correct HRGs improved when coders used either case notes or medical support in addition to the discharge summary. Further emphasis needs to be placed on improving the standard of information recorded in discharge summaries. Copyright © 2018 Elsevier B.V. All rights reserved.

  16. A method for modeling co-occurrence propensity of clinical codes with application to ICD-10-PCS auto-coding.

    Science.gov (United States)

    Subotin, Michael; Davis, Anthony R

    2016-09-01

    Natural language processing methods for medical auto-coding, or automatic generation of medical billing codes from electronic health records, generally assign each code independently of the others. They may thus assign codes for closely related procedures or diagnoses to the same document, even when they do not tend to occur together in practice, simply because the right choice can be difficult to infer from the clinical narrative. We propose a method that injects awareness of the propensities for code co-occurrence into this process. First, a model is trained to estimate the conditional probability that one code is assigned by a human coder, given than another code is known to have been assigned to the same document. Then, at runtime, an iterative algorithm is used to apply this model to the output of an existing statistical auto-coder to modify the confidence scores of the codes. We tested this method in combination with a primary auto-coder for International Statistical Classification of Diseases-10 procedure codes, achieving a 12% relative improvement in F-score over the primary auto-coder baseline. The proposed method can be used, with appropriate features, in combination with any auto-coder that generates codes with different levels of confidence. The promising results obtained for International Statistical Classification of Diseases-10 procedure codes suggest that the proposed method may have wider applications in auto-coding. © The Author 2016. Published by Oxford University Press on behalf of the American Medical Informatics Association. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  17. Application of Monte Carlo method in study of the padronization for radionuclides with complex disintegration scheme in 4{pi}{beta}-{gamma} coincidence System; Aplicacao do metodo de Monte Carlo no estudo da padronizacao de radionuclideos com esquema de desintegracao complexos em sistema de coincidencias 4{pi}{beta}-{gamma}

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Mauro Noriaki

    2006-07-01

    The present work described a new methodology for modelling the behaviour of the activity in a 4{pi}{beta}-{gamma} coincidence system. The detection efficiency for electrons in the proportional counter and gamma radiation in the NaI(Tl) detector was calculated using the Monte Carlo program MCNP4C. Another Monte Carlo code was developed which follows the path in the disintegration scheme from the initial state of the precursor radionuclide, until the ground state of the daughter nucleus. Every step of the disintegration scheme is sorted by random numbers taking into account the probabilities of all {beta}{sup -} branches, electronic capture branches, transitions probabilities and internal conversion coefficients. Once the final state was reached beta, electronic capture events and gamma transitions are accounted for the three spectra: beta, gamma and coincidence variation in the beta efficiency was performed simulating energy cut off or use of absorbers (Collodion). The selected radionuclides for simulation were: {sup 134}Cs, {sup 72}Ga which disintegrate by {beta}{sup -} transition, {sup 133}Ba which disintegrates by electronic capture and {sup 35}S which is a beta pure emitter. For the latter, the Efficiency Tracing technique was simulated. The extrapolation curves obtained by Monte Carlo were filled by the Least Square Method with the experimental points and the results were compared to the Linear Extrapolation method. (author)

  18. A prospect for the development of an epithermal neutron beam from the horizontal channel at the TRNC for brain tumors treatment based on the BNCT method

    International Nuclear Information System (INIS)

    Ben-Ghazail, Mustafa Ali

    2005-01-01

    In this work the epithermal neutron was development from horizontal channel VI at Tajoura research reactor which can be used for Boron Neutron Capture Therapy. The analysis of reactivity and control rod worth is performed by three dimensional continues energy MCNP-4C code with neutron cross section data from the ENDF/B-VI evaluation. The neutron beam which is developed for medical purpose is generated from the reactor core by means of U-235 fission. The neutrons leaking through the cavity of HC in Be-9 reflector is guided through a tube made of stainless steel to patient position. The HC has two wheels. The first wheel is small and is used as a gate. The second is large and have three positions one to close the gate, the second to open the gate while the third for loading collimator. The collimator consists of the moderators and filters to optimize the neutron beam which is installed in the loading position. The HC VI is extended to the room constructed to allow space for other horizontal channels users. materials are used to optimize the neutron beam which was selected depending on neutron beam properties related to core loading and control rod position. The results of the development study show that the required values for the neutron beam characteristic can be nearly reached. The different comparisons of the calculations performed using MCNP-4C code with the requirements values of characteristics neutron beam show that the result values of MCNP-4C code model are reliable. (author)

  19. Local coding based matching kernel method for image classification.

    Directory of Open Access Journals (Sweden)

    Yan Song

    Full Text Available This paper mainly focuses on how to effectively and efficiently measure visual similarity for local feature based representation. Among existing methods, metrics based on Bag of Visual Word (BoV techniques are efficient and conceptually simple, at the expense of effectiveness. By contrast, kernel based metrics are more effective, but at the cost of greater computational complexity and increased storage requirements. We show that a unified visual matching framework can be developed to encompass both BoV and kernel based metrics, in which local kernel plays an important role between feature pairs or between features and their reconstruction. Generally, local kernels are defined using Euclidean distance or its derivatives, based either explicitly or implicitly on an assumption of Gaussian noise. However, local features such as SIFT and HoG often follow a heavy-tailed distribution which tends to undermine the motivation behind Euclidean metrics. Motivated by recent advances in feature coding techniques, a novel efficient local coding based matching kernel (LCMK method is proposed. This exploits the manifold structures in Hilbert space derived from local kernels. The proposed method combines advantages of both BoV and kernel based metrics, and achieves a linear computational complexity. This enables efficient and scalable visual matching to be performed on large scale image sets. To evaluate the effectiveness of the proposed LCMK method, we conduct extensive experiments with widely used benchmark datasets, including 15-Scenes, Caltech101/256, PASCAL VOC 2007 and 2011 datasets. Experimental results confirm the effectiveness of the relatively efficient LCMK method.

  20. Resonance interference method in lattice physics code stream

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

    2015-01-01

    Newly developed resonance interference model is implemented in the lattice physics code STREAM, and the model shows a significant improvement in computing accurate eigenvalues. Equivalence theory is widely used in production calculations to generate the effective multigroup (MG) cross-sections (XS) for commercial reactors. Although a lot of methods have been developed to enhance the accuracy in computing effective XSs, the current resonance treatment methods still do not have a clear resonance interference model. The conventional resonance interference model simply adds the absorption XSs of resonance isotopes to the background XS. However, the conventional models show non-negligible errors in computing effective XSs and eigenvalues. In this paper, a resonance interference factor (RIF) library method is proposed. This method interpolates the RIFs in a pre-generated RIF library and corrects the effective XS, rather than solving the time consuming slowing down calculation. The RIF library method is verified for homogeneous and heterogeneous problems. The verification results using the proposed method show significant improvements of accuracy in treating the interference effect. (author)

  1. A gridding method for object-oriented PIC codes

    International Nuclear Information System (INIS)

    Gisler, G.; Peter, W.; Nash, H.; Acquah, J.; Lin, C.; Rine, D.

    1993-01-01

    A simple, rule-based gridding method for object-oriented PIC codes is described which is not only capable of dealing with complicated structures such as multiply-connected regions, but is also computationally faster than classical gridding techniques. Using, these smart grids, vacant cells (e.g., cells enclosed by conductors) will never have to be stored or calculated, thus avoiding the usual situation of having to zero electromagnetic fields within conductors after valuable cpu time has been spent in calculating the fields within these cells in the first place. This object-oriented gridding technique makes use of encapsulating characteristics of actual physical objects (particles, fields, grids, etc.) in C ++ classes and supporting software reuse of these entities through C ++ class inheritance relations. It has been implemented in the form of a simple two-dimensional plasma particle-in-cell code, and forms the initial effort of an AFOSR research project to develop a flexible software simulation environment for particle-in-cell algorithms based on object-oriented technology

  2. Present status of transport code development based on Monte Carlo method

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki

    1985-01-01

    The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)

  3. The comparison of MCNP perturbation technique with MCNP difference method in critical calculation

    International Nuclear Information System (INIS)

    Liu Bin; Lv Xuefeng; Zhao Wei; Wang Kai; Tu Jing; Ouyang Xiaoping

    2010-01-01

    For a nuclear fission system, we calculated Δk eff , which arise from system material composition changes, by two different approaches, the MCNP perturbation technique and the MCNP difference method. For every material composition change, we made four different runs, each run with different cycles or each cycle generating different neutrons, then we compared the two Δk eff that are obtained by two different approaches. As a material composition change in any particular cell of the nuclear fission system is small compared to the material compositions in the whole nuclear fission system, in other words, this composition change can be treated as a small perturbation, the Δk eff results obtained from the MCNP perturbation technique are much quicker, much more efficient and reliable than the results from the MCNP difference method. When a material composition change in any particular cell of the nuclear fission system is significant compared to the material compositions in the whole nuclear fission system, both the MCNP perturbation technique and the MCNP difference method can give satisfactory results. But for the run with the same cycles and each cycle generating the same neutrons, the results obtained from the MCNP perturbation technique are systemically less than the results obtained from the MCNP difference method. To further confirm our calculation results from the MCNP4C, we run the exact same MCNP4C input file in MCNP5, the calculation results from MCNP5 are the same as the calculation results from MCNP4C. We need caution when using the MCNP perturbation technique to calculate the Δk eff as the material composition change is large compared to the material compositions in the whole nuclear fission system, even though the material composition changes of any particular cell of the fission system still meet the criteria of MCNP perturbation technique.

  4. A new method for species identification via protein-coding and non-coding DNA barcodes by combining machine learning with bioinformatic methods.

    Directory of Open Access Journals (Sweden)

    Ai-bing Zhang

    Full Text Available Species identification via DNA barcodes is contributing greatly to current bioinventory efforts. The initial, and widely accepted, proposal was to use the protein-coding cytochrome c oxidase subunit I (COI region as the standard barcode for animals, but recently non-coding internal transcribed spacer (ITS genes have been proposed as candidate barcodes for both animals and plants. However, achieving a robust alignment for non-coding regions can be problematic. Here we propose two new methods (DV-RBF and FJ-RBF to address this issue for species assignment by both coding and non-coding sequences that take advantage of the power of machine learning and bioinformatics. We demonstrate the value of the new methods with four empirical datasets, two representing typical protein-coding COI barcode datasets (neotropical bats and marine fish and two representing non-coding ITS barcodes (rust fungi and brown algae. Using two random sub-sampling approaches, we demonstrate that the new methods significantly outperformed existing Neighbor-joining (NJ and Maximum likelihood (ML methods for both coding and non-coding barcodes when there was complete species coverage in the reference dataset. The new methods also out-performed NJ and ML methods for non-coding sequences in circumstances of potentially incomplete species coverage, although then the NJ and ML methods performed slightly better than the new methods for protein-coding barcodes. A 100% success rate of species identification was achieved with the two new methods for 4,122 bat queries and 5,134 fish queries using COI barcodes, with 95% confidence intervals (CI of 99.75-100%. The new methods also obtained a 96.29% success rate (95%CI: 91.62-98.40% for 484 rust fungi queries and a 98.50% success rate (95%CI: 96.60-99.37% for 1094 brown algae queries, both using ITS barcodes.

  5. A new method for species identification via protein-coding and non-coding DNA barcodes by combining machine learning with bioinformatic methods.

    Science.gov (United States)

    Zhang, Ai-bing; Feng, Jie; Ward, Robert D; Wan, Ping; Gao, Qiang; Wu, Jun; Zhao, Wei-zhong

    2012-01-01

    Species identification via DNA barcodes is contributing greatly to current bioinventory efforts. The initial, and widely accepted, proposal was to use the protein-coding cytochrome c oxidase subunit I (COI) region as the standard barcode for animals, but recently non-coding internal transcribed spacer (ITS) genes have been proposed as candidate barcodes for both animals and plants. However, achieving a robust alignment for non-coding regions can be problematic. Here we propose two new methods (DV-RBF and FJ-RBF) to address this issue for species assignment by both coding and non-coding sequences that take advantage of the power of machine learning and bioinformatics. We demonstrate the value of the new methods with four empirical datasets, two representing typical protein-coding COI barcode datasets (neotropical bats and marine fish) and two representing non-coding ITS barcodes (rust fungi and brown algae). Using two random sub-sampling approaches, we demonstrate that the new methods significantly outperformed existing Neighbor-joining (NJ) and Maximum likelihood (ML) methods for both coding and non-coding barcodes when there was complete species coverage in the reference dataset. The new methods also out-performed NJ and ML methods for non-coding sequences in circumstances of potentially incomplete species coverage, although then the NJ and ML methods performed slightly better than the new methods for protein-coding barcodes. A 100% success rate of species identification was achieved with the two new methods for 4,122 bat queries and 5,134 fish queries using COI barcodes, with 95% confidence intervals (CI) of 99.75-100%. The new methods also obtained a 96.29% success rate (95%CI: 91.62-98.40%) for 484 rust fungi queries and a 98.50% success rate (95%CI: 96.60-99.37%) for 1094 brown algae queries, both using ITS barcodes.

  6. A Semantic Analysis Method for Scientific and Engineering Code

    Science.gov (United States)

    Stewart, Mark E. M.

    1998-01-01

    This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

  7. Step by step parallel programming method for molecular dynamics code

    International Nuclear Information System (INIS)

    Orii, Shigeo; Ohta, Toshio

    1996-07-01

    Parallel programming for a numerical simulation program of molecular dynamics is carried out with a step-by-step programming technique using the two phase method. As a result, within the range of a certain computing parameters, it is found to obtain parallel performance by using the level of parallel programming which decomposes the calculation according to indices of do-loops into each processor on the vector parallel computer VPP500 and the scalar parallel computer Paragon. It is also found that VPP500 shows parallel performance in wider range computing parameters. The reason is that the time cost of the program parts, which can not be reduced by the do-loop level of the parallel programming, can be reduced to the negligible level by the vectorization. After that, the time consuming parts of the program are concentrated on less parts that can be accelerated by the do-loop level of the parallel programming. This report shows the step-by-step parallel programming method and the parallel performance of the molecular dynamics code on VPP500 and Paragon. (author)

  8. Application of the NJOY code for unresolved resonance treatment in the MCNP utility code

    International Nuclear Information System (INIS)

    Milosevic, M.; Greenspan, E.; Vujic, J. . E-mail addresses of corresponding authors: mmilos@vin.bg.ac.yu , vujic@nuc.berkeley.edu ,; Milosevic, M.; Vujic, J.)

    2005-01-01

    There are numerous uncertainties in the prediction of neutronic characteristics of reactor cores, particularly in the case of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) benchmark core and the new procedures and cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and it is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include: MOCUP - a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the ENDF/B-VI evaluations; and KWO2 - a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 group library. Calculations made for the ENHS benchmark have shown that the differences between the results obtained using different code systems and cross section libraries are significant and should be taken into account in assessing the quality of nuclear data libraries. (author)

  9. Recent advances in neutral particle transport methods and codes

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1996-01-01

    An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned

  10. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    International Nuclear Information System (INIS)

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

  11. Optimized Method for Generating and Acquiring GPS Gold Codes

    Directory of Open Access Journals (Sweden)

    Khaled Rouabah

    2015-01-01

    Full Text Available We propose a simpler and faster Gold codes generator, which can be efficiently initialized to any desired code, with a minimum delay. Its principle consists of generating only one sequence (code number 1 from which we can produce all the other different signal codes. This is realized by simply shifting this sequence by different delays that are judiciously determined by using the bicorrelation function characteristics. This is in contrast to the classical Linear Feedback Shift Register (LFSR based Gold codes generator that requires, in addition to the shift process, a significant number of logic XOR gates and a phase selector to change the code. The presence of all these logic XOR gates in classical LFSR based Gold codes generator provokes the consumption of an additional time in the generation and acquisition processes. In addition to its simplicity and its rapidity, the proposed architecture, due to the total absence of XOR gates, has fewer resources than the conventional Gold generator and can thus be produced at lower cost. The Digital Signal Processing (DSP implementations have shown that the proposed architecture presents a solution for acquiring Global Positioning System (GPS satellites signals optimally and in a parallel way.

  12. A novel construction method of QC-LDPC codes based on CRT for optical communications

    Science.gov (United States)

    Yuan, Jian-guo; Liang, Meng-qi; Wang, Yong; Lin, Jin-zhao; Pang, Yu

    2016-05-01

    A novel construction method of quasi-cyclic low-density parity-check (QC-LDPC) codes is proposed based on Chinese remainder theory (CRT). The method can not only increase the code length without reducing the girth, but also greatly enhance the code rate, so it is easy to construct a high-rate code. The simulation results show that at the bit error rate ( BER) of 10-7, the net coding gain ( NCG) of the regular QC-LDPC(4 851, 4 546) code is respectively 2.06 dB, 1.36 dB, 0.53 dB and 0.31 dB more than those of the classic RS(255, 239) code in ITU-T G.975, the LDPC(32 640, 30 592) code in ITU-T G.975.1, the QC-LDPC(3 664, 3 436) code constructed by the improved combining construction method based on CRT and the irregular QC-LDPC(3 843, 3 603) code constructed by the construction method based on the Galois field ( GF( q)) multiplicative group. Furthermore, all these five codes have the same code rate of 0.937. Therefore, the regular QC-LDPC(4 851, 4 546) code constructed by the proposed construction method has excellent error-correction performance, and can be more suitable for optical transmission systems.

  13. Improved Intra-coding Methods for H.264/AVC

    Directory of Open Access Journals (Sweden)

    Li Song

    2009-01-01

    Full Text Available The H.264/AVC design adopts a multidirectional spatial prediction model to reduce spatial redundancy, where neighboring pixels are used as a prediction for the samples in a data block to be encoded. In this paper, a recursive prediction scheme and an enhanced (block-matching algorithm BMA prediction scheme are designed and integrated into the state-of-the-art H.264/AVC framework to provide a new intra coding model. Extensive experiments demonstrate that the coding efficiency can be on average increased by 0.27 dB with comparison to the performance of the conventional H.264 coding model.

  14. An Efficient Method for Verifying Gyrokinetic Microstability Codes

    Science.gov (United States)

    Bravenec, R.; Candy, J.; Dorland, W.; Holland, C.

    2009-11-01

    Benchmarks for gyrokinetic microstability codes can be developed through successful ``apples-to-apples'' comparisons among them. Unlike previous efforts, we perform the comparisons for actual discharges, rendering the verification efforts relevant to existing experiments and future devices (ITER). The process requires i) assembling the experimental analyses at multiple times, radii, discharges, and devices, ii) creating the input files ensuring that the input parameters are faithfully translated code-to-code, iii) running the codes, and iv) comparing the results, all in an organized fashion. The purpose of this work is to automate this process as much as possible: At present, a python routine is used to generate and organize GYRO input files from TRANSP or ONETWO analyses. Another routine translates the GYRO input files into GS2 input files. (Translation software for other codes has not yet been written.) Other python codes submit the multiple GYRO and GS2 jobs, organize the results, and collect them into a table suitable for plotting. (These separate python routines could easily be consolidated.) An example of the process -- a linear comparison between GYRO and GS2 for a DIII-D discharge at multiple radii -- will be presented.

  15. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  16. Fluid dynamics and heat transfer methods for the TRAC code

    International Nuclear Information System (INIS)

    Reed, W.H.; Kirchner, W.L.

    1977-01-01

    A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, we have developed new highly implicit difference techniques that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained. (author)

  17. A robust fusion method for multiview distributed video coding

    DEFF Research Database (Denmark)

    Salmistraro, Matteo; Ascenso, Joao; Brites, Catarina

    2014-01-01

    Distributed video coding (DVC) is a coding paradigm which exploits the redundancy of the source (video) at the decoder side, as opposed to predictive coding, where the encoder leverages the redundancy. To exploit the correlation between views, multiview predictive video codecs require the encoder...... with a robust fusion system able to improve the quality of the fused SI along the decoding process through a learning process using already decoded data. We shall here take the approach to fuse the estimated distributions of the SIs as opposed to a conventional fusion algorithm based on the fusion of pixel...... values. The proposed solution is able to achieve gains up to 0.9 dB in Bjøntegaard difference when compared with the best-performing (in a RD sense) single SI DVC decoder, chosen as the best of an inter-view and a temporal SI-based decoder one....

  18. Fluid dynamics and heat transfer methods for the TRAC code

    International Nuclear Information System (INIS)

    Reed, W.H.; Kirchner, W.L.

    1977-01-01

    A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, new highly implicit difference techniques are developed that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained

  19. Improving the safety of a body composition analyser based on the PGNAA method

    Energy Technology Data Exchange (ETDEWEB)

    Miri-Hakimabad, Hashem; Izadi-Najafabadi, Reza; Vejdani-Noghreiyan, Alireza; Panjeh, Hamed [FUM Radiation Detection And Measurement Laboratory, Ferdowsi University of Mashhad (Iran, Islamic Republic of)

    2007-12-15

    The {sup 252}Cf radioisotope and {sup 241}Am-Be are intense neutron emitters that are readily encapsulated in compact, portable and sealed sources. Some features such as high flux of neutron emission and reliable neutron spectrum of these sources make them suitable for the prompt gamma neutron activation analysis (PGNAA) method. The PGNAA method can be used in medicine for neutron radiography and body chemical composition analysis. {sup 252}Cf and {sup 241}Am-Be sources generate not only neutrons but also are intense gamma emitters. Furthermore, the sample in medical treatments is a human body, so it may be exposed to the bombardments of these gamma-rays. Moreover, accumulations of these high-rate gamma-rays in the detector volume cause simultaneous pulses that can be piled up and distort the spectra in the region of interest (ROI). In order to remove these disadvantages in a practical way without being concerned about losing the thermal neutron flux, a gamma-ray filter made of Pb must be employed. The paper suggests a relatively safe body chemical composition analyser (BCCA) machine that uses a spherical Pb shield, enclosing the neutron source. Gamma-ray shielding effects and the optimum radius of the spherical Pb shield have been investigated, using the MCNP-4C code, and compared with the unfiltered case, the bare source. Finally, experimental results demonstrate that an optimised gamma-ray shield for the neutron source in a BCCA can reduce effectively the risk of exposure to the {sup 252}Cf and {sup 241}Am-Be sources.

  20. Mapping Saldana's Coding Methods onto the Literature Review Process

    Science.gov (United States)

    Onwuegbuzie, Anthony J.; Frels, Rebecca K.; Hwang, Eunjin

    2016-01-01

    Onwuegbuzie and Frels (2014) provided a step-by-step guide illustrating how discourse analysis can be used to analyze literature. However, more works of this type are needed to address the way that counselor researchers conduct literature reviews. Therefore, we present a typology for coding and analyzing information extracted for literature…

  1. Method for quantitative assessment of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Dearien, J.A.; Davis, C.B.; Matthews, L.J.

    1979-01-01

    A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

  2. TASS/SMR Code Topical Report for SMART Plant, Vol. I: Code Structure, System Models, and Solution Methods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)

    2008-10-15

    The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.

  3. Methods for the development of large computer codes under LTSS

    International Nuclear Information System (INIS)

    Sicilian, J.M.

    1977-06-01

    TRAC is a large computer code being developed by Group Q-6 for the analysis of the transient thermal hydraulic behavior of light-water nuclear reactors. A system designed to assist the development of TRAC is described. The system consists of a central HYDRA dataset, R6LIB, containing files used in the development of TRAC, and a file maintenance program, HORSE, which facilitates the use of this dataset

  4. Advancing methods for reliably assessing motivational interviewing fidelity using the motivational interviewing skills code.

    Science.gov (United States)

    Lord, Sarah Peregrine; Can, Doğan; Yi, Michael; Marin, Rebeca; Dunn, Christopher W; Imel, Zac E; Georgiou, Panayiotis; Narayanan, Shrikanth; Steyvers, Mark; Atkins, David C

    2015-02-01

    The current paper presents novel methods for collecting MISC data and accurately assessing reliability of behavior codes at the level of the utterance. The MISC 2.1 was used to rate MI interviews from five randomized trials targeting alcohol and drug use. Sessions were coded at the utterance-level. Utterance-based coding reliability was estimated using three methods and compared to traditional reliability estimates of session tallies. Session-level reliability was generally higher compared to reliability using utterance-based codes, suggesting that typical methods for MISC reliability may be biased. These novel methods in MI fidelity data collection and reliability assessment provided rich data for therapist feedback and further analyses. Beyond implications for fidelity coding, utterance-level coding schemes may elucidate important elements in the counselor-client interaction that could inform theories of change and the practice of MI. Copyright © 2015 Elsevier Inc. All rights reserved.

  5. CREPT-MCNP code for efficiency calibration of HPGe detectors with the representative point method.

    Science.gov (United States)

    Saegusa, Jun

    2008-01-01

    The representative point method for the efficiency calibration of volume samples has been previously proposed. For smoothly implementing the method, a calculation code named CREPT-MCNP has been developed. The code estimates the position of a representative point which is intrinsic to each shape of volume sample. The self-absorption correction factors are also given to make correction on the efficiencies measured at the representative point with a standard point source. Features of the CREPT-MCNP code are presented.

  6. Development of 'SKYSHINE-CG' code. A line-beam method code equipped with combinatorial geometry routine

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Takahiro; Ochiai, Katsuharu [Plant and System Planning Department, Toshiba Corporation, Yokohama, Kanagawa (Japan); Uematsu, Mikio; Hayashida, Yoshihisa [Department of Nuclear Engineering, Toshiba Engineering Corporation, Yokohama, Kanagawa (Japan)

    2000-03-01

    A boiling water reactor (BWR) plant has a single loop coolant system, in which main steam generated in the reactor core proceeds directly into turbines. Consequently, radioactive {sup 16}N (6.2 MeV photon emitter) contained in the steam contributes to gamma-ray skyshine dose in the vicinity of the BWR plant. The skyshine dose analysis is generally performed with the line-beam method code SKYSHINE, in which calculational geometry consists of a rectangular turbine building and a set of isotropic point sources corresponding to an actual distribution of {sup 16}N sources. For the purpose of upgrading calculational accuracy, the SKYSHINE-CG code has been developed by incorporating the combinatorial geometry (CG) routine into the SKYSHINE code, so that shielding effect of in-building equipment can be properly considered using a three-dimensional model composed of boxes, cylinders, spheres, etc. Skyshine dose rate around a 500 MWe BWR plant was calculated with both SKYSHINE and SKYSHINE-CG codes, and the calculated results were compared with measured data obtained with a NaI(Tl) scintillation detector. The C/E values for SKYSHINE-CG calculation were scattered around 4.0, whereas the ones for SKYSHINE calculation were as large as 6.0. Calculational error was found to be reduced by adopting three-dimensional model based on the combinatorial geometry method. (author)

  7. Development of methodology for characterization of cartridge filters from the IEA-R1 using the Monte Carlo method

    International Nuclear Information System (INIS)

    Costa, Priscila

    2014-01-01

    The Cuno filter is part of the water processing circuit of the IEA-R1 reactor and, when saturated, it is replaced and becomes a radioactive waste, which must be managed. In this work, the primary characterization of the Cuno filter of the IEA-R1 nuclear reactor at IPEN was carried out using gamma spectrometry associated with the Monte Carlo method. The gamma spectrometry was performed using a hyperpure germanium detector (HPGe). The germanium crystal represents the detection active volume of the HPGe detector, which has a region called dead layer or inactive layer. It has been reported in the literature a difference between the theoretical and experimental values when obtaining the efficiency curve of these detectors. In this study we used the MCNP-4C code to obtain the detector calibration efficiency for the geometry of the Cuno filter, and the influence of the dead layer and the effect of sum in cascade at the HPGe detector were studied. The correction of the dead layer values were made by varying the thickness and the radius of the germanium crystal. The detector has 75.83 cm 3 of active volume of detection, according to information provided by the manufacturer. Nevertheless, the results showed that the actual value of active volume is less than the one specified, where the dead layer represents 16% of the total volume of the crystal. A Cuno filter analysis by gamma spectrometry has enabled identifying energy peaks. Using these peaks, three radionuclides were identified in the filter: 108m Ag, 110m Ag and 60 Co. From the calibration efficiency obtained by the Monte Carlo method, the value of activity estimated for these radionuclides is in the order of MBq. (author)

  8. Methods of evaluating the effects of coding on SAR data

    Science.gov (United States)

    Dutkiewicz, Melanie; Cumming, Ian

    1993-01-01

    It is recognized that mean square error (MSE) is not a sufficient criterion for determining the acceptability of an image reconstructed from data that has been compressed and decompressed using an encoding algorithm. In the case of Synthetic Aperture Radar (SAR) data, it is also deemed to be insufficient to display the reconstructed image (and perhaps error image) alongside the original and make a (subjective) judgment as to the quality of the reconstructed data. In this paper we suggest a number of additional evaluation criteria which we feel should be included as evaluation metrics in SAR data encoding experiments. These criteria have been specifically chosen to provide a means of ensuring that the important information in the SAR data is preserved. The paper also presents the results of an investigation into the effects of coding on SAR data fidelity when the coding is applied in (1) the signal data domain, and (2) the image domain. An analysis of the results highlights the shortcomings of the MSE criterion, and shows which of the suggested additional criterion have been found to be most important.

  9. CodeRAnts: A recommendation method based on collaborative searching and ant colonies, applied to reusing of open source code

    Directory of Open Access Journals (Sweden)

    Isaac Caicedo-Castro

    2014-01-01

    Full Text Available This paper presents CodeRAnts, a new recommendation method based on a collaborative searching technique and inspired on the ant colony metaphor. This method aims to fill the gap in the current state of the matter regarding recommender systems for software reuse, for which prior works present two problems. The first is that, recommender systems based on these works cannot learn from the collaboration of programmers and second, outcomes of assessments carried out on these systems present low precision measures and recall and in some of these systems, these metrics have not been evaluated. The work presented in this paper contributes a recommendation method, which solves these problems.

  10. A method for generating subgroup parameters from resonance tables and the SPART code

    International Nuclear Information System (INIS)

    Devan, K.; Mohanakrishnan, P.

    1995-01-01

    A method for generating subgroup or band parameters from resonance tables is described. A computer code SPART was written using this method. This code generates the subgroup parameters for any number of bands within the specified broad groups at different temperatures by reading the required input data from the binary cross section library in the Cadarache format. The results obtained with SPART code for two bands were compared with that obtained from GROUPIE code and a good agreement was obtained. Results of the generation of subgroup parameters in four bands for sample case of 239 Pu from resonance tables of Cadarache Ver.2 library is also presented. 6 refs, 2 tabs

  11. A method for scientific code coupling in a distributed environment; Une methodologie pour le couplage de codes scientifiques en environnement distribue

    Energy Technology Data Exchange (ETDEWEB)

    Caremoli, C; Beaucourt, D; Chen, O; Nicolas, G; Peniguel, C; Rascle, P; Richard, N; Thai Van, D; Yessayan, A

    1994-12-01

    This guide book deals with coupling of big scientific codes. First, the context is introduced: big scientific codes devoted to a specific discipline coming to maturity, and more and more needs in terms of multi discipline studies. Then we describe different kinds of code coupling and an example of code coupling: 3D thermal-hydraulic code THYC and 3D neutronics code COCCINELLE. With this example we identify problems to be solved to realize a coupling. We present the different numerical methods usable for the resolution of coupling terms. This leads to define two kinds of coupling: with the leak coupling, we can use explicit methods, and with the strong coupling we need to use implicit methods. On both cases, we analyze the link with the way of parallelizing code. For translation of data from one code to another, we define the notion of Standard Coupling Interface based on a general structure for data. This general structure constitutes an intermediary between the codes, thus allowing a relative independence of the codes from a specific coupling. The proposed method for the implementation of a coupling leads to a simultaneous run of the different codes, while they exchange data. Two kinds of data communication with message exchange are proposed: direct communication between codes with the use of PVM product (Parallel Virtual Machine) and indirect communication with a coupling tool. This second way, with a general code coupling tool, is based on a coupling method, and we strongly recommended to use it. This method is based on the two following principles: re-usability, that means few modifications on existing codes, and definition of a code usable for coupling, that leads to separate the design of a code usable for coupling from the realization of a specific coupling. This coupling tool available from beginning of 1994 is described in general terms. (authors). figs., tabs.

  12. MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System

    International Nuclear Information System (INIS)

    2007-01-01

    A - Description of program or function: MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC, nor from the NEA Databank. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included. The following changes have been made: 1) An increase in the number of removal group information that must be provided for each material in each step in the feed input file. 2) The capability to use CINDER90 instead of ORIGEN2.1 as the depletion/decay part of the code. 3) ORIGEN2.2 can also be used instead of ORIGEN2.1 in Monteburns. 4) The correction of including the capture cross sections to metastable as well as ground states if applicable for an isotope (i.e. Am-241 and Am-243 in particular). 5) The ability to use a MCNP input file that has a title card starting with 'm' (this was a bug in the first version of Monteburns). 6) A decrease in run time for cases involving decay-only steps (power of 0.0). Monteburns does not run MCNP to calculate cross sections for a step unless it is an irradiation step. 7) The ability to change the cross section libraries used each step. If different cross section libraries are desired for multiple steps. 8

  13. Coupling of partitioned physics codes with quasi-Newton methods

    CSIR Research Space (South Africa)

    Haelterman, R

    2017-03-01

    Full Text Available , A class of methods for solving nonlinear simultaneous equations. Math. Comp. 19, pp. 577–593 (1965) [3] C.G. Broyden, Quasi-Newton methods and their applications to function minimization. Math. Comp. 21, pp. 368–381 (1967) [4] J.E. Dennis, J.J. More...´, Quasi-Newton methods: motivation and theory. SIAM Rev. 19, pp. 46–89 (1977) [5] J.E. Dennis, R.B. Schnabel, Least Change Secant Updates for quasi- Newton methods. SIAM Rev. 21, pp. 443–459 (1979) [6] G. Dhondt, CalculiX CrunchiX USER’S MANUAL Version 2...

  14. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  15. WKB: an interactive code for solving differential equations using phase integral methods

    International Nuclear Information System (INIS)

    White, R.B.

    1978-01-01

    A small code for the analysis of ordinary differential equations interactively through the use of Phase Integral Methods (WKB) has been written for use on the DEC 10. This note is a descriptive manual for those interested in using the code

  16. Development of three-dimensional transport code by the double finite element method

    International Nuclear Information System (INIS)

    Fujimura, Toichiro

    1985-01-01

    Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)

  17. Double folding model of nucleus-nucleus potential: formulae, iteration method and computer code

    International Nuclear Information System (INIS)

    Luk'yanov, K.V.

    2008-01-01

    Method of construction of the nucleus-nucleus double folding potential is described. Iteration procedure for the corresponding integral equation is presented. Computer code and numerical results are presented

  18. Calculating Error Percentage in Using Water Phantom Instead of Soft Tissue Concerning 103Pd Brachytherapy Source Distribution via Monte Carlo Method

    Directory of Open Access Journals (Sweden)

    OL Ahmadi

    2015-12-01

    Full Text Available Introduction: 103Pd is a low energy source, which is used in brachytherapy. According to the standards of American Association of Physicists in Medicine, dosimetric parameters determination of brachytherapy sources before the clinical application was considered significantly important. Therfore, the present study aimed to compare the dosimetric parameters of the target source using the water phantom and soft tissue. Methods: According to the TG-43U1 protocol, the dosimetric parameters were compared around the 103Pd source in regard with water phantom with the density of 0.998 gr/cm3 and the soft tissue with the density of 1.04 gr/cm3 on the longitudinal and transverse axes using the MCNP4C code and the relative differences were compared between the both conditions. Results: The simulation results indicated that the dosimetric parameters depended on the radial dose function and the anisotropy function in the application of the water phantom instead of soft tissue up to a distance of 1.5 cm,  between which a good consistency was observed. With increasing the distance, the difference increased, so as within 6 cm from the source, this difference increased to 4%. Conclusions: The results of  the soft tissue phantom compared with those of the water phantom indicated 4% relative difference at a distance of 6 cm from the source. Therefore, the results of the water phantom with a maximum error of 4% can be used in practical applications instead of soft tissue. Moreover, the amount of differences obtained in each distance regarding using the soft tissue phantom could be corrected.

  19. Compatibility of global environmental assessment methods of buildings with an Egyptian energy code

    Directory of Open Access Journals (Sweden)

    Amal Kamal Mohamed Shamseldin

    2017-04-01

    Full Text Available Several environmental assessment methods of buildings had emerged over the world to set environmental classifications for buildings, such as the American method “Leadership in Energy and Environmental Design” (LEED the most widespread one. Several countries decided to put their own assessment methods to catch up with the previous orientation, such as Egypt. The main goal of putting the Egyptian method was to impose the voluntary local energy efficiency codes. Through a local survey, it was clearly noted that many of the construction makers in Egypt do not even know the local method, and whom are interested in the environmental assessment of buildings seek to apply LEED rather than anything else. Therefore, several questions appear about the American method compatibility with the Egyptian energy codes – that contain the most exact characteristics and requirements and give the outmost credible energy efficiency results for buildings in Egypt-, and the possibility of finding another global method that gives closer results to those of the Egyptian codes, especially with the great variety of energy efficiency measurement approaches used among the different assessment methods. So, the researcher is trying to find the compatibility of using non-local assessment methods with the local energy efficiency codes. Thus, if the results are not compatible, the Egyptian government should take several steps to increase the local building sector awareness of the Egyptian method to benefit these codes, and it should begin to enforce it within the building permits after a proper guidance and feedback.

  20. How recalibration method, pricing, and coding affect DRG weights

    Science.gov (United States)

    Carter, Grace M.; Rogowski, Jeannette A.

    1992-01-01

    We compared diagnosis-related group (DRG) weights calculated using the hospital-specific relative-value (HSR V) methodology with those calculated using the standard methodology for each year from 1985 through 1989 and analyzed differences between the two methods in detail for 1989. We provide evidence suggesting that classification error and subsidies of higher weighted cases by lower weighted cases caused compression in the weights used for payment as late as the fifth year of the prospective payment system. However, later weights calculated by the standard method are not compressed because a statistical correlation between high markups and high case-mix indexes offsets the cross-subsidization. HSR V weights from the same files are compressed because this methodology is more sensitive to cross-subsidies. However, both sets of weights produce equally good estimates of hospital-level costs net of those expenses that are paid by outlier payments. The greater compression of the HSR V weights is counterbalanced by the fact that more high-weight cases qualify as outliers. PMID:10127456

  1. Development of methodology for characterization of cartridge filters from the IEA-R1 using the Monte Carlo method; Desenvolvimento de uma metodologia para caracterizacao do filtro cuno do reator IEA-R1 utilizando o Metodo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Priscila

    2014-07-01

    The Cuno filter is part of the water processing circuit of the IEA-R1 reactor and, when saturated, it is replaced and becomes a radioactive waste, which must be managed. In this work, the primary characterization of the Cuno filter of the IEA-R1 nuclear reactor at IPEN was carried out using gamma spectrometry associated with the Monte Carlo method. The gamma spectrometry was performed using a hyperpure germanium detector (HPGe). The germanium crystal represents the detection active volume of the HPGe detector, which has a region called dead layer or inactive layer. It has been reported in the literature a difference between the theoretical and experimental values when obtaining the efficiency curve of these detectors. In this study we used the MCNP-4C code to obtain the detector calibration efficiency for the geometry of the Cuno filter, and the influence of the dead layer and the effect of sum in cascade at the HPGe detector were studied. The correction of the dead layer values were made by varying the thickness and the radius of the germanium crystal. The detector has 75.83 cm{sup 3} of active volume of detection, according to information provided by the manufacturer. Nevertheless, the results showed that the actual value of active volume is less than the one specified, where the dead layer represents 16% of the total volume of the crystal. A Cuno filter analysis by gamma spectrometry has enabled identifying energy peaks. Using these peaks, three radionuclides were identified in the filter: {sup 108m}Ag, {sup 110m}Ag and {sup 60}Co. From the calibration efficiency obtained by the Monte Carlo method, the value of activity estimated for these radionuclides is in the order of MBq. (author)

  2. Development of burnup methods and capabilities in Monte Carlo code RMC

    International Nuclear Information System (INIS)

    She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

    2013-01-01

    Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

  3. Status of SFR Codes and Methods QA Implementation

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States); Briggs, Laural L. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, Thomas H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-31

    This report details development of the SAS4A/SASSYS-1 SQA Program and describes the initial stages of Program implementation planning. The provisional Program structure, which is largely focused on the establishment of compliant SQA documentation, is outlined in detail, and Program compliance with the appropriate SQA requirements is highlighted. Additional program activities, such as improvements to testing methods and Program surveillance, are also described in this report. Given that the programmatic resources currently granted to development of the SAS4A/SASSYS-1 SQA Program framework are not sufficient to adequately address all SQA requirements (e.g. NQA-1, NUREG/BR-0167, etc.), this report also provides an overview of the gaps that remain the SQA program, and highlights recommendations on a path forward to resolution of these issues. One key finding of this effort is the identification of the need for an SQA program sustainable over multiple years within DOE annual R&D funding constraints.

  4. Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.; Gvildys, J.

    1977-01-01

    In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs

  5. An Efficient Integer Coding and Computing Method for Multiscale Time Segment

    Directory of Open Access Journals (Sweden)

    TONG Xiaochong

    2016-12-01

    Full Text Available This article focus on the exist problem and status of current time segment coding, proposed a new set of approach about time segment coding: multi-scale time segment integer coding (MTSIC. This approach utilized the tree structure and the sort by size formed among integer, it reflected the relationship among the multi-scale time segments: order, include/contained, intersection, etc., and finally achieved an unity integer coding processing for multi-scale time. On this foundation, this research also studied the computing method for calculating the time relationships of MTSIC, to support an efficient calculation and query based on the time segment, and preliminary discussed the application method and prospect of MTSIC. The test indicated that, the implement of MTSIC is convenient and reliable, and the transformation between it and the traditional method is convenient, it has the very high efficiency in query and calculating.

  6. Refuelling design and core calculations at NPP Paks: codes and methods

    International Nuclear Information System (INIS)

    Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

    2001-01-01

    This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

  7. A generic method for automatic translation between input models for different versions of simulation codes

    International Nuclear Information System (INIS)

    Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

    2014-01-01

    A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications

  8. A generic method for automatic translation between input models for different versions of simulation codes

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School of Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

    2014-05-01

    A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications.

  9. Licensing in BE system code calculations. Applications and uncertainty evaluation by CIAU method

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco

    2007-01-01

    The evaluation of uncertainty constitutes the necessary supplement of Best Estimate (BE) calculations performed to understand accident scenarios in water cooled nuclear reactors. The needs come from the imperfection of computational tools on the one side and from the interest in using such tool to get more precise evaluation of safety margins. In the present paper the approaches to uncertainty are outlined and the CIAU (Code with capability of Internal Assessment of Uncertainty) method proposed by the University of Pisa is described including ideas at the basis and results from applications. Two approaches are distinguished that are characterized as 'propagation of code input uncertainty' and 'propagation of code output errors'. For both methods, the thermal-hydraulic code is at the centre of the process of uncertainty evaluation: in the former case the code itself is adopted to compute the error bands and to propagate the input errors, in the latter case the errors in code application to relevant measurements are used to derive the error bands. The CIAU method exploits the idea of the 'status approach' for identifying the thermal-hydraulic conditions of an accident in any Nuclear Power Plant (NPP). Errors in predicting such status are derived from the comparison between predicted and measured quantities and, in the stage of the application of the method, are used to compute the uncertainty. (author)

  10. SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON

    International Nuclear Information System (INIS)

    BEEBE - WANG, J.; LUCCIO, A.U.; D IMPERIO, N.; MACHIDA, S.

    2002-01-01

    Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed

  11. SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON.

    Energy Technology Data Exchange (ETDEWEB)

    BEEBE - WANG,J.; LUCCIO,A.U.; D IMPERIO,N.; MACHIDA,S.

    2002-06-03

    Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed.

  12. Average Likelihood Methods of Classification of Code Division Multiple Access (CDMA)

    Science.gov (United States)

    2016-05-01

    subject to code matrices that follows the structure given by (113). [⃗ yR y⃗I ] = √ Es 2L [ GR1 −GI1 GI2 GR2 ] [ QR −QI QI QR ] [⃗ bR b⃗I ] + [⃗ nR n⃗I... QR ] [⃗ b+ b⃗− ] + [⃗ n+ n⃗− ] (115) The average likelihood for type 4 CDMA (116) is a special case of type 1 CDMA with twice the code length and...AVERAGE LIKELIHOOD METHODS OF CLASSIFICATION OF CODE DIVISION MULTIPLE ACCESS (CDMA) MAY 2016 FINAL TECHNICAL REPORT APPROVED FOR PUBLIC RELEASE

  13. An improved method for storing and retrieving tabulated data in a scalar Monte Carlo code

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Reynolds, K.H.; Dodds, H.L.; Landers, N.F.; Petrie, L.M.

    1990-01-01

    The KENO-Va code is a production-level criticality safety code used to calculate the k eff of a system. The code is stochastic in nature, using a Monte Carlo algorithm to track individual particles one at a time through the system. The advent of computers with vector processors has generated an interest in improving KENO-Va to take advantage of the potential speed-up associated with these new processors. Unfortunately, the original Monte Carlo algorithm and method of storing and retrieving cross-section data is not adaptable to vector processing. This paper discusses an alternate method for storing and retrieving data that not only is readily vectorizable but also improves the efficiency of the current scalar code

  14. A New Image Encryption Technique Combining Hill Cipher Method, Morse Code and Least Significant Bit Algorithm

    Science.gov (United States)

    Nofriansyah, Dicky; Defit, Sarjon; Nurcahyo, Gunadi W.; Ganefri, G.; Ridwan, R.; Saleh Ahmar, Ansari; Rahim, Robbi

    2018-01-01

    Cybercrime is one of the most serious threats. Efforts are made to reduce the number of cybercrime is to find new techniques in securing data such as Cryptography, Steganography and Watermarking combination. Cryptography and Steganography is a growing data security science. A combination of Cryptography and Steganography is one effort to improve data integrity. New techniques are used by combining several algorithms, one of which is the incorporation of hill cipher method and Morse code. Morse code is one of the communication codes used in the Scouting field. This code consists of dots and lines. This is a new modern and classic concept to maintain data integrity. The result of the combination of these three methods is expected to generate new algorithms to improve the security of the data, especially images.

  15. Coupling methods for parallel running RELAPSim codes in nuclear power plant simulation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yankai; Lin, Meng, E-mail: linmeng@sjtu.edu.cn; Yang, Yanhua

    2016-02-15

    When the plant is modeled detailedly for high precision, it is hard to achieve real-time calculation for one single RELAP5 in a large-scale simulation. To improve the speed and ensure the precision of simulation at the same time, coupling methods for parallel running RELAPSim codes were proposed in this study. Explicit coupling method via coupling boundaries was realized based on a data-exchange and procedure-control environment. Compromise of synchronization frequency was well considered to improve the precision of simulation and guarantee the real-time simulation at the same time. The coupling methods were assessed using both single-phase flow models and two-phase flow models and good agreements were obtained between the splitting–coupling models and the integrated model. The mitigation of SGTR was performed as an integral application of the coupling models. A large-scope NPP simulator was developed adopting six splitting–coupling models of RELAPSim and other simulation codes. The coupling models could improve the speed of simulation significantly and make it possible for real-time calculation. In this paper, the coupling of the models in the engineering simulator is taken as an example to expound the coupling methods, i.e., coupling between parallel running RELAPSim codes, and coupling between RELAPSim code and other types of simulation codes. However, the coupling methods are also referable in other simulator, for example, a simulator employing ATHLETE instead of RELAP5, other logic code instead of SIMULINK. It is believed the coupling method is commonly used for NPP simulator regardless of the specific codes chosen in this paper.

  16. Method for calculating internal radiation and ventilation with the ADINAT heat-flow code

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Montan, D.N.

    1980-01-01

    One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation and ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation

  17. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.; De Crecy, F.

    1994-01-01

    This paper deals with the application of the adjoint sensitivity method (ASM) to thermal hydraulic codes. The advantage of the method is to use small central processing unit time in comparison with the usual approach requiring one complete code run per sensitivity determination. In the first part the mathematical aspects of the problem are treated, and the applicability of the method of the functional-type response of a thermal hydraulic model is demonstrated. On a simple example of non-linear hyperbolic equation (Burgers equation) the problem has been analysed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the continuous ASM and the discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the discrete ASM constitutes a practical solution for thermal hydraulic codes. The application of the discrete ASM to the thermal hydraulic safety code CATHARE is then presented for two examples. They demonstrate that the discrete ASM constitutes an efficient tool for the analysis of code sensitivity. ((orig.))

  18. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Crecy, F. de; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  19. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  20. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  1. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Crecy, F de; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  2. Analysis of the applicability of acceleration methods for a triangular prism geometry nodal diffusion code

    International Nuclear Information System (INIS)

    Fujimura, Toichiro; Okumura, Keisuke

    2002-11-01

    A prototype version of a diffusion code has been developed to analyze the hexagonal core as reduced moderation reactor and the applicability of some acceleration methods have been investigated to accelerate the convergence of the iterative solution method. The hexagonal core is divided into regular triangular prisms in the three-dimensional code MOSRA-Prism and a polynomial expansion nodal method is applied to approximate the neutron flux distribution by a cubic polynomial. The multi-group diffusion equation is solved iteratively with ordinal inner and outer iterations and the effectiveness of acceleration methods is ascertained by applying an adaptive acceleration method and a neutron source extrapolation method, respectively. The formulation of the polynomial expansion nodal method is outlined in the report and the local and global effectiveness of the acceleration methods is discussed with various sample calculations. A new general expression of vacuum boundary condition, derived in the formulation is also described. (author)

  3. Hybrid Micro-Depletion method in the DYN3D code

    Energy Technology Data Exchange (ETDEWEB)

    Bilodid, Yurii [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

    2016-07-01

    A new method for accounting spectral history effects was developed and implemented in the reactor dynamics code DYN3D. Detailed nuclide content is calculated for each region of the reactor core and used to correct fuel properties. The new method demonstrates excellent results in test cases.

  4. Beacon- and Schema-Based Method for Recognizing Algorithms from Students' Source Code

    Science.gov (United States)

    Taherkhani, Ahmad; Malmi, Lauri

    2013-01-01

    In this paper, we present a method for recognizing algorithms from students programming submissions coded in Java. The method is based on the concept of "programming schemas" and "beacons". Schemas are high-level programming knowledge with detailed knowledge abstracted out, and beacons are statements that imply specific…

  5. Application of Wielandt method in continuous-energy nuclear data sensitivity analysis with RMC code

    International Nuclear Information System (INIS)

    Qiu Yishu; Wang Kan; She Ding

    2015-01-01

    The Iterated Fission Probability (IFP) method, an accurate method to estimate adjoint-weighted quantities in the continuous-energy Monte Carlo criticality calculations, has been widely used for calculating kinetic parameters and nuclear data sensitivity coefficients. By using a strategy of waiting, however, this method faces the challenge of high memory usage to store the tallies of original contributions which size is proportional to the number of particle histories in each cycle. Recently, the Wielandt method, applied by Monte Carlo code McCARD to calculate kinetic parameters, estimates adjoint fluxes in a single particle history and thus can save memory usage. In this work, the Wielandt method has been applied in Rector Monte Carlo code RMC for nuclear data sensitivity analysis. The methodology and algorithm of applying Wielandt method in estimation of adjoint-based sensitivity coefficients are discussed. Verification is performed by comparing the sensitivity coefficients calculated by Wielandt method with analytical solutions, those computed by IFP method which is also implemented in RMC code for sensitivity analysis, and those from the multi-group TSUNAMI-3D module in SCALE code package. (author)

  6. The OpenMOC method of characteristics neutral particle transport code

    International Nuclear Information System (INIS)

    Boyd, William; Shaner, Samuel; Li, Lulu; Forget, Benoit; Smith, Kord

    2014-01-01

    Highlights: • An open source method of characteristics neutron transport code has been developed. • OpenMOC shows nearly perfect scaling on CPUs and 30× speedup on GPUs. • Nonlinear acceleration techniques demonstrate a 40× reduction in source iterations. • OpenMOC uses modern software design principles within a C++ and Python framework. • Validation with respect to the C5G7 and LRA benchmarks is presented. - Abstract: The method of characteristics (MOC) is a numerical integration technique for partial differential equations, and has seen widespread use for reactor physics lattice calculations. The exponential growth in computing power has finally brought the possibility for high-fidelity full core MOC calculations within reach. The OpenMOC code is being developed at the Massachusetts Institute of Technology to investigate algorithmic acceleration techniques and parallel algorithms for MOC. OpenMOC is a free, open source code written using modern software languages such as C/C++ and CUDA with an emphasis on extensible design principles for code developers and an easy to use Python interface for code users. The present work describes the OpenMOC code and illustrates its ability to model large problems accurately and efficiently

  7. Modeling radiation belt dynamics using a 3-D layer method code

    Science.gov (United States)

    Wang, C.; Ma, Q.; Tao, X.; Zhang, Y.; Teng, S.; Albert, J. M.; Chan, A. A.; Li, W.; Ni, B.; Lu, Q.; Wang, S.

    2017-08-01

    A new 3-D diffusion code using a recently published layer method has been developed to analyze radiation belt electron dynamics. The code guarantees the positivity of the solution even when mixed diffusion terms are included. Unlike most of the previous codes, our 3-D code is developed directly in equatorial pitch angle (α0), momentum (p), and L shell coordinates; this eliminates the need to transform back and forth between (α0,p) coordinates and adiabatic invariant coordinates. Using (α0,p,L) is also convenient for direct comparison with satellite data. The new code has been validated by various numerical tests, and we apply the 3-D code to model the rapid electron flux enhancement following the geomagnetic storm on 17 March 2013, which is one of the Geospace Environment Modeling Focus Group challenge events. An event-specific global chorus wave model, an AL-dependent statistical plasmaspheric hiss wave model, and a recently published radial diffusion coefficient formula from Time History of Events and Macroscale Interactions during Substorms (THEMIS) statistics are used. The simulation results show good agreement with satellite observations, in general, supporting the scenario that the rapid enhancement of radiation belt electron flux for this event results from an increased level of the seed population by radial diffusion, with subsequent acceleration by chorus waves. Our results prove that the layer method can be readily used to model global radiation belt dynamics in three dimensions.

  8. MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) code description and User's Manual

    International Nuclear Information System (INIS)

    Avci, H.I.; Raghuram, S.; Baybutt, P.

    1985-04-01

    A new computer code called MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) has been developed to replace the CORRAL-2 computer code which was written for the Reactor Safety Study (WASH-1400). This report is a User's Manual for MATADOR. MATADOR is intended for use in system risk studies to analyze radionuclide transport and deposition in reactor containments. The principal output of the code is information on the timing and magnitude of radionuclide releases to the environment as a result of severely degraded core accidents. MATADOR considers the transport of radionuclides through the containment and their removal by natural deposition and by engineered safety systems such as sprays. It is capable of analyzing the behavior of radionuclides existing either as vapors or aerosols in the containment. The code requires input data on the source terms into the containment, the geometry of the containment, and thermal-hydraulic conditions in the containment

  9. WASTK: A Weighted Abstract Syntax Tree Kernel Method for Source Code Plagiarism Detection

    Directory of Open Access Journals (Sweden)

    Deqiang Fu

    2017-01-01

    Full Text Available In this paper, we introduce a source code plagiarism detection method, named WASTK (Weighted Abstract Syntax Tree Kernel, for computer science education. Different from other plagiarism detection methods, WASTK takes some aspects other than the similarity between programs into account. WASTK firstly transfers the source code of a program to an abstract syntax tree and then gets the similarity by calculating the tree kernel of two abstract syntax trees. To avoid misjudgment caused by trivial code snippets or frameworks given by instructors, an idea similar to TF-IDF (Term Frequency-Inverse Document Frequency in the field of information retrieval is applied. Each node in an abstract syntax tree is assigned a weight by TF-IDF. WASTK is evaluated on different datasets and, as a result, performs much better than other popular methods like Sim and JPlag.

  10. Development of a CAD-based neutron transport code with the method of characteristics

    International Nuclear Information System (INIS)

    Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing

    2012-01-01

    The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)

  11. Methods for Coding Tobacco-Related Twitter Data: A Systematic Review.

    Science.gov (United States)

    Lienemann, Brianna A; Unger, Jennifer B; Cruz, Tess Boley; Chu, Kar-Hai

    2017-03-31

    As Twitter has grown in popularity to 313 million monthly active users, researchers have increasingly been using it as a data source for tobacco-related research. The objective of this systematic review was to assess the methodological approaches of categorically coded tobacco Twitter data and make recommendations for future studies. Data sources included PsycINFO, Web of Science, PubMed, ABI/INFORM, Communication Source, and Tobacco Regulatory Science. Searches were limited to peer-reviewed journals and conference proceedings in English from January 2006 to July 2016. The initial search identified 274 articles using a Twitter keyword and a tobacco keyword. One coder reviewed all abstracts and identified 27 articles that met the following inclusion criteria: (1) original research, (2) focused on tobacco or a tobacco product, (3) analyzed Twitter data, and (4) coded Twitter data categorically. One coder extracted data collection and coding methods. E-cigarettes were the most common type of Twitter data analyzed, followed by specific tobacco campaigns. The most prevalent data sources were Gnip and Twitter's Streaming application programming interface (API). The primary methods of coding were hand-coding and machine learning. The studies predominantly coded for relevance, sentiment, theme, user or account, and location of user. Standards for data collection and coding should be developed to be able to more easily compare and replicate tobacco-related Twitter results. Additional recommendations include the following: sample Twitter's databases multiple times, make a distinction between message attitude and emotional tone for sentiment, code images and URLs, and analyze user profiles. Being relatively novel and widely used among adolescents and black and Hispanic individuals, Twitter could provide a rich source of tobacco surveillance data among vulnerable populations. ©Brianna A Lienemann, Jennifer B Unger, Tess Boley Cruz, Kar-Hai Chu. Originally published in the

  12. Introduction into scientific work methods-a necessity when performance-based codes are introduced

    DEFF Research Database (Denmark)

    Dederichs, Anne; Sørensen, Lars Schiøtt

    The introduction of performance-based codes in Denmark in 2004 requires new competences from people working with different aspects of fire safety in the industry and the public sector. This abstract presents an attempt in reducing problems with handling and analysing the mathematical methods...... and CFD models when applying performance-based codes. This is done within the educational program "Master of Fire Safety Engineering" at the department of Civil Engineering at the Technical University of Denmark. It was found that the students had general problems with academic methods. Therefore, a new...

  13. Efficient depth intraprediction method for H.264/AVC-based three-dimensional video coding

    Science.gov (United States)

    Oh, Kwan-Jung; Oh, Byung Tae

    2015-04-01

    We present an intracoding method that is applicable to depth map coding in multiview plus depth systems. Our approach combines skip prediction and plane segmentation-based prediction. The proposed depth intraskip prediction uses the estimated direction at both the encoder and decoder, and does not need to encode residual data. Our plane segmentation-based intraprediction divides the current block into biregions, and applies a different prediction scheme for each segmented region. This method avoids incorrect estimations across different regions, resulting in higher prediction accuracy. Simulation results demonstrate that the proposed scheme is superior to H.264/advanced video coding intraprediction and has the ability to improve the subjective rendering quality.

  14. Computer-modeling codes to improve exploration nuclear-logging methods. National Uranium Resource Evaluation

    International Nuclear Information System (INIS)

    Wilson, R.D.; Price, R.K.; Kosanke, K.L.

    1983-03-01

    As part of the Department of Energy's National Uranium Resource Evaluation (NURE) project's Technology Development effort, a number of computer codes and accompanying data bases were assembled for use in modeling responses of nuclear borehole logging Sondes. The logging methods include fission neutron, active and passive gamma-ray, and gamma-gamma. These CDC-compatible computer codes and data bases are available on magnetic tape from the DOE Technical Library at its Grand Junction Area Office. Some of the computer codes are standard radiation-transport programs that have been available to the radiation shielding community for several years. Other codes were specifically written to model the response of borehole radiation detectors or are specialized borehole modeling versions of existing Monte Carlo transport programs. Results from several radiation modeling studies are available as two large data bases (neutron and gamma-ray). These data bases are accompanied by appropriate processing programs that permit the user to model a wide range of borehole and formation-parameter combinations for fission-neutron, neutron-, activation and gamma-gamma logs. The first part of this report consists of a brief abstract for each code or data base. The abstract gives the code name and title, short description, auxiliary requirements, typical running time (CDC 6600), and a list of references. The next section gives format specifications and/or directory for the tapes. The final section of the report presents listings for programs used to convert data bases between machine floating-point and EBCDIC

  15. Computerization of reporting and data storage using automatic coding method in the department of radiology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Hee; Lee, Kyung Sang; Kim, Woo Ho; Han, Joon Koo; Choi, Byung Ihn; Han, Man Chung [College of Medicine, Seoul National Univ., Seoul (Korea, Republic of)

    1990-10-15

    The authors developed a computer program for use in printing report as well as data storage and retrieval in the Radiology department. This program used IBM PC AT and was written in dBASE III plus language. The automatic coding method of the ACR code, developed by Kim et al was applied in this program, and the framework of this program is the same as that developed for the surgical pathology department. The working sheet, which contained the name card for X-ray film identification and the results of previous radiologic studies, were printed during registration. The word precessing function was applied for issuing the formal report of radiologic study, and the data storage was carried out during the typewriting of the report. Two kinds of data files were stored in the hard disk ; the temporary file contained full information and the permanent file contained patient's identification data, and ACR code. Searching of a specific case was performed by chart number, patients name, date of study, or ACR code within a second. All the cases were arranged by ACR codes of procedure code, anatomy code, and pathology code. Every new data was copied to the diskette after daily work automatically, with which data could be restored in case of hard diskette failure. The main advantage of this program with comparison to the larger computer system is its low price. Based on the experience in the Seoul District Armed Forces General Hospital, we assume that this program provides solution to various problems in the radiology department where a large computer system with well designed software is not available.

  16. GPU-accelerated 3D neutron diffusion code based on finite difference method

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Q.; Yu, G.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ. (China)

    2012-07-01

    Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

  17. GPU-accelerated 3D neutron diffusion code based on finite difference method

    International Nuclear Information System (INIS)

    Xu, Q.; Yu, G.; Wang, K.

    2012-01-01

    Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

  18. Comparison of different methods used in integral codes to model coagulation of aerosols

    Science.gov (United States)

    Beketov, A. I.; Sorokin, A. A.; Alipchenkov, V. M.; Mosunova, N. A.

    2013-09-01

    The methods for calculating coagulation of particles in the carrying phase that are used in the integral codes SOCRAT, ASTEC, and MELCOR, as well as the Hounslow and Jacobson methods used to model aerosol processes in the chemical industry and in atmospheric investigations are compared on test problems and against experimental results in terms of their effectiveness and accuracy. It is shown that all methods are characterized by a significant error in modeling the distribution function for micrometer particles if calculations are performed using rather "coarse" spectra of particle sizes, namely, when the ratio of the volumes of particles from neighboring fractions is equal to or greater than two. With reference to the problems considered, the Hounslow method and the method applied in the aerosol module used in the ASTEC code are the most efficient ones for carrying out calculations.

  19. Implementation of the dynamic Monte Carlo method for transient analysis in the general purpose code Tripoli

    Energy Technology Data Exchange (ETDEWEB)

    Sjenitzer, Bart L.; Hoogenboom, J. Eduard, E-mail: B.L.Sjenitzer@TUDelft.nl, E-mail: J.E.Hoogenboom@TUDelft.nl [Delft University of Technology (Netherlands)

    2011-07-01

    A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)

  20. Implementation of the dynamic Monte Carlo method for transient analysis in the general purpose code Tripoli

    International Nuclear Information System (INIS)

    Sjenitzer, Bart L.; Hoogenboom, J. Eduard

    2011-01-01

    A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)

  1. Packing simulation code to calculate distribution function of hard spheres by Monte Carlo method : MCRDF

    International Nuclear Information System (INIS)

    Murata, Isao; Mori, Takamasa; Nakagawa, Masayuki; Shirai, Hiroshi.

    1996-03-01

    High Temperature Gas-cooled Reactors (HTGRs) employ spherical fuels named coated fuel particles (CFPs) consisting of a microsphere of low enriched UO 2 with coating layers in order to prevent FP release. There exist many spherical fuels distributed randomly in the cores. Therefore, the nuclear design of HTGRs is generally performed on the basis of the multigroup approximation using a diffusion code, S N transport code or group-wise Monte Carlo code. This report summarizes a Monte Carlo hard sphere packing simulation code to simulate the packing of equal hard spheres and evaluate the necessary probability distribution of them, which is used for the application of the new Monte Carlo calculation method developed to treat randomly distributed spherical fuels with the continuous energy Monte Carlo method. By using this code, obtained are the various statistical values, namely Radial Distribution Function (RDF), Nearest Neighbor Distribution (NND), 2-dimensional RDF and so on, for random packing as well as ordered close packing of FCC and BCC. (author)

  2. Stochastic methods for uncertainty treatment of functional variables in computer codes: application to safety studies

    International Nuclear Information System (INIS)

    Nanty, Simon

    2015-01-01

    This work relates to the framework of uncertainty quantification for numerical simulators, and more precisely studies two industrial applications linked to the safety studies of nuclear plants. These two applications have several common features. The first one is that the computer code inputs are functional and scalar variables, functional ones being dependent. The second feature is that the probability distribution of functional variables is known only through a sample of their realizations. The third feature, relative to only one of the two applications, is the high computational cost of the code, which limits the number of possible simulations. The main objective of this work was to propose a complete methodology for the uncertainty analysis of numerical simulators for the two considered cases. First, we have proposed a methodology to quantify the uncertainties of dependent functional random variables from a sample of their realizations. This methodology enables to both model the dependency between variables and their link to another variable, called co-variate, which could be, for instance, the output of the considered code. Then, we have developed an adaptation of a visualization tool for functional data, which enables to simultaneously visualize the uncertainties and features of dependent functional variables. Second, a method to perform the global sensitivity analysis of the codes used in the two studied cases has been proposed. In the case of a computationally demanding code, the direct use of quantitative global sensitivity analysis methods is intractable. To overcome this issue, the retained solution consists in building a surrogate model or meta model, a fast-running model approximating the computationally expensive code. An optimized uniform sampling strategy for scalar and functional variables has been developed to build a learning basis for the meta model. Finally, a new approximation approach for expensive codes with functional outputs has been

  3. Building America Guidance for Identifying and Overcoming Code, Standard, and Rating Method Barriers

    Energy Technology Data Exchange (ETDEWEB)

    Cole, P. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Halverson, M. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-01

    This guidance document was prepared using the input from the meeting summarized in the draft CSI Roadmap to provide Building America research teams and partners with specific information and approaches to identifying and overcoming potential barriers to Building America innovations arising in and/or stemming from codes, standards, and rating methods.

  4. The Effects of Single and Dual Coded Multimedia Instructional Methods on Chinese Character Learning

    Science.gov (United States)

    Wang, Ling

    2013-01-01

    Learning Chinese characters is a difficult task for adult English native speakers due to the significant differences between the Chinese and English writing system. The visuospatial properties of Chinese characters have inspired the development of instructional methods using both verbal and visual information based on the Dual Coding Theory. This…

  5. Method and device for fast code acquisition in spread spectrum receivers

    NARCIS (Netherlands)

    Coenen, A.J.R.M.

    1993-01-01

    Abstract of NL 9101155 (A) Method for code acquisition in a satellite receiver. The biphase-modulated high-frequency carrier transmitted by a satellite is converted via a fixed local oscillator frequency down to the baseband, whereafter the baseband signal is fed via a bandpass filter, which has an

  6. SQA of finite element method (FEM) codes used for analyses of pit storage/transport packages

    Energy Technology Data Exchange (ETDEWEB)

    Russel, E. [Lawrence Livermore National Lab., CA (United States)

    1997-11-01

    This report contains viewgraphs on the software quality assurance of finite element method codes used for analyses of pit storage and transport projects. This methodology utilizes the ISO 9000-3: Guideline for application of 9001 to the development, supply, and maintenance of software, for establishing well-defined software engineering processes to consistently maintain high quality management approaches.

  7. Implantation of a new calculation method of fuel depletion in the CITHAM code

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    It is evaluated the accuracy of the linear aproximation method used in the CITHAN code to obtain the solution of depletion equations. Results are compared with the Benchmark problem. The convenience of depletion chain before criticality calculations is analysed. The depletion calculation was modified using linear combination technic of linear chains. (M.C.K.) [pt

  8. Fuel penetration of intersubassembly gaps in LMFBRs: a calculational method with the SIMMER-II code

    International Nuclear Information System (INIS)

    DeVault, G.P.

    1983-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor (LMFBR) undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. A possible avenue for early fuel removal in heterogeneous core LMFBRs is the failure of duct walls in disrupted driver subassemblies followed by fuel penetration into the gaps between blanket subassemblies. The SIMMER-II code was modified to simulate flow between subassembly gaps. Calculations with the modified SIMMER-II code indicate the capabilities of the method and the potential for fuel mass reduction in the active core

  9. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1995-01-01

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  10. TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

    International Nuclear Information System (INIS)

    Shen, H.; Li, Z.; Wang, K.; Yu, G.

    2010-01-01

    A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

  11. Probability-neighbor method of accelerating geometry treatment in reactor Monte Carlo code RMC

    International Nuclear Information System (INIS)

    She, Ding; Li, Zeguang; Xu, Qi; Wang, Kan; Yu, Ganglin

    2011-01-01

    Probability neighbor method (PNM) is proposed in this paper to accelerate geometry treatment of Monte Carlo (MC) simulation and validated in self-developed reactor Monte Carlo code RMC. During MC simulation by either ray-tracking or delta-tracking method, large amounts of time are spent in finding out which cell one particle is located in. The traditional way is to search cells one by one with certain sequence defined previously. However, this procedure becomes very time-consuming when the system contains a large number of cells. Considering that particles have different probability to enter different cells, PNM method optimizes the searching sequence, i.e., the cells with larger probability are searched preferentially. The PNM method is implemented in RMC code and the numerical results show that the considerable time of geometry treatment in MC calculation for complicated systems is saved, especially effective in delta-tracking simulation. (author)

  12. Computing eigenvalue sensitivity coefficients to nuclear data based on the CLUTCH method with RMC code

    International Nuclear Information System (INIS)

    Qiu, Yishu; She, Ding; Tang, Xiao; Wang, Kan; Liang, Jingang

    2016-01-01

    Highlights: • A new algorithm is proposed to reduce memory consumption for sensitivity analysis. • The fission matrix method is used to generate adjoint fission source distributions. • Sensitivity analysis is performed on a detailed 3D full-core benchmark with RMC. - Abstract: Recently, there is a need to develop advanced methods of computing eigenvalue sensitivity coefficients to nuclear data in the continuous-energy Monte Carlo codes. One of these methods is the iterated fission probability (IFP) method, which is adopted by most of Monte Carlo codes of having the capabilities of computing sensitivity coefficients, including the Reactor Monte Carlo code RMC. Though it is accurate theoretically, the IFP method faces the challenge of huge memory consumption. Therefore, it may sometimes produce poor sensitivity coefficients since the number of particles in each active cycle is not sufficient enough due to the limitation of computer memory capacity. In this work, two algorithms of the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method, namely, the collision-event-based algorithm (C-CLUTCH) which is also implemented in SCALE and the fission-event-based algorithm (F-CLUTCH) which is put forward in this work, are investigated and implemented in RMC to reduce memory requirements for computing eigenvalue sensitivity coefficients. While the C-CLUTCH algorithm requires to store concerning reaction rates of every collision, the F-CLUTCH algorithm only stores concerning reaction rates of every fission point. In addition, the fission matrix method is put forward to generate the adjoint fission source distribution for the CLUTCH method to compute sensitivity coefficients. These newly proposed approaches implemented in RMC code are verified by a SF96 lattice model and the MIT BEAVRS benchmark problem. The numerical results indicate the accuracy of the F-CLUTCH algorithm is the same as the C

  13. Structural dynamics in LMFBR containment analysis. A brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers to up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. First, the commonly used finite-difference expressions in the Lagrangian codes will be compared. Sample calculations will be used as a basis for discussing and comparing the accuracy of the various finite-difference representations. The distortion of the meshes will also be compared; the techniques used for eliminating the numerical instabilities will be discussed and compared using examples. Next, the numerical methods used in the Eulerian formulation will be compared, first among themselves and then with the Lagrangian formulations. Special emphasis is placed on the effect of mass diffusion of the Eulerian calculation on the propagation of discontinuities. Implicit and explicit numerical integrations will be discussed and results obtained from these two techniques will be compared. Then, the finite-element methods are compared with the finite-difference methods. The advantages and disadvantages of the two methods will be discussed in detail, together with the versatility and ease of application of the method to containment analysis having complex geometries. It will also be shown that the finite-element equations for a constant-pressure fluid element is identical to the finite-difference equations using contour integrations. Finally, conclusions based on this study will be given

  14. Comparisons of coded aperture imaging using various apertures and decoding methods

    International Nuclear Information System (INIS)

    Chang, L.T.; Macdonald, B.; Perez-Mendez, V.

    1976-07-01

    The utility of coded aperture γ camera imaging of radioisotope distributions in Nuclear Medicine is in its ability to give depth information about a three dimensional source. We have calculated imaging with Fresnel zone plate and multiple pinhole apertures to produce coded shadows and reconstruction of these shadows using correlation, Fresnel diffraction, and Fourier transform deconvolution. Comparisons of the coded apertures and decoding methods are made by evaluating their point response functions both for in-focus and out-of-focus image planes. Background averages and standard deviations were calculated. In some cases, background subtraction was made using combinations of two complementary apertures. Results using deconvolution reconstruction for finite numbers of events are also given

  15. GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications

    International Nuclear Information System (INIS)

    Glaeser, H.

    2008-01-01

    During the recent years, an increasing interest in computational reactor safety analysis is to replace the conservative evaluation model calculations by best estimate calculations supplemented by uncertainty analysis of the code results. The evaluation of the margin to acceptance criteria, for example, the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. Uncertainty analysis is needed if useful conclusions are to be obtained from best estimate thermal-hydraulic code calculations, otherwise single values of unknown accuracy would be presented for comparison with regulatory acceptance limits. Methods have been developed and presented to quantify the uncertainty of computer code results. The basic techniques proposed by GRS are presented together with applications to a large break loss of coolant accident on a reference reactor as well as on an experiment simulating containment behaviour

  16. Building America Guidance for Identifying and Overcoming Code, Standard, and Rating Method Barriers

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Pamala C.; Halverson, Mark A.

    2013-09-01

    The U.S. Department of Energy’s (DOE) Building America program implemented a new Codes and Standards Innovation (CSI) Team in 2013. The Team’s mission is to assist Building America (BA) research teams and partners in identifying and resolving conflicts between Building America innovations and the various codes and standards that govern the construction of residences. A CSI Roadmap was completed in September, 2013. This guidance document was prepared using the information in the CSI Roadmap to provide BA research teams and partners with specific information and approaches to identifying and overcoming potential barriers to Building America (BA) innovations arising in and/or stemming from codes, standards, and rating methods. For more information on the BA CSI team, please email: CSITeam@pnnl.gov

  17. An efficient simulation method of a cyclotron sector-focusing magnet using 2D Poisson code

    Energy Technology Data Exchange (ETDEWEB)

    Gad Elmowla, Khaled Mohamed M; Chai, Jong Seo, E-mail: jschai@skku.edu; Yeon, Yeong H; Kim, Sangbum; Ghergherehchi, Mitra

    2016-10-01

    In this paper we discuss design simulations of a spiral magnet using 2D Poisson code. The Independent Layers Method (ILM) is a new technique that was developed to enable the use of two-dimensional simulation code to calculate a non-symmetric 3-dimensional magnetic field. In ILM, the magnet pole is divided into successive independent layers, and the hill and valley shape around the azimuthal direction is implemented using a reference magnet. The normalization of the magnetic field in the reference magnet produces a profile that can be multiplied by the maximum magnetic field in the hill magnet, which is a dipole magnet made of the hills at the same radius. Both magnets are then calculated using the 2D Poisson SUPERFISH code. Then a fully three-dimensional magnetic field is produced using TOSCA for the original spiral magnet, and the comparison of the 2D and 3D results shows a good agreement between both.

  18. Development of improved methods for the LWR lattice physics code EPRI-CELL

    International Nuclear Information System (INIS)

    Williams, M.L.; Wright, R.Q.; Barhen, J.

    1982-07-01

    A number of improvements have been made by ORNL to the lattice physics code EPRI-CELL (E-C) which is widely used by utilities for analysis of power reactors. The code modifications were made mainly in the thermal and epithermal routines and resulted in improved reactor physics approximations and more efficient running times. The improvements in the thermal flux calculation included implementation of a group-dependent rebalance procedure to accelerate the iterative process and a more rigorous calculation of interval-to-interval collision probabilities. The epithermal resonance shielding methods used in the code have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology

  19. Coding "We-ness" in couple's relationship stories: A method for assessing mutuality in couple therapy.

    Science.gov (United States)

    Gildersleeve, Sara; Singer, Jefferson A; Skerrett, Karen; Wein, Shelter

    2017-05-01

    "We-ness," a couple's mutual investment in their relationship and in each other, has been found to be a potent dimension of couple resilience. This study examined the development of a method to capture We-ness in psychotherapy through the coding of relationship narratives co-constructed by couples ("We-Stories"). It used a coding system to identify the core thematic elements that make up these narratives. Couples that self-identified as "happy" (N = 53) generated We-Stories and completed measures of relationship satisfaction and mutuality. These stories were then coded using the We-Stories coding manual. Findings indicated that security, an element that involves aspects of safety, support, and commitment, was most common, appearing in 58.5% of all narratives. This element was followed by the elements of pleasure (49.1%) and shared meaning/vision (37.7%). The number of "We-ness" elements was also correlated with and predictive of discrepancy scores on measures of relationship mutuality, indicating the validity of the We-Stories coding manual. Limitations and future directions are discussed.

  20. Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

    The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

  1. 3-D spherical harmonics code FFT3 by the finite Fourier transformation method

    International Nuclear Information System (INIS)

    Kobayashi, K.

    1997-01-01

    In the odd order spherical harmonics method, the rigorous boundary condition at the material interfaces is that the even moments of the angular flux and the normal components of the even order moments of current vectors must be continuous. However, it is difficult to derive spatial discretized equations by the finite difference or finite element methods, which satisfy this material interface condition. It is shown that using the finite Fourier transformation method, space discretized equations which satisfy this interface condition can be easily derived. The discrepancies of the flux distribution near void region between spherical harmonics method codes may be due to the difference of application of the material interface condition. (author)

  2. A perturbation-based susbtep method for coupled depletion Monte-Carlo codes

    International Nuclear Information System (INIS)

    Kotlyar, Dan; Aufiero, Manuele; Shwageraus, Eugene; Fratoni, Massimiliano

    2017-01-01

    Highlights: • The GPT method allows to calculate the sensitivity coefficients to any perturbation. • Full Jacobian of sensitivities, cross sections (XS) to concentrations, may be obtained. • The time dependent XS is obtained by combining the GPT and substep methods. • The proposed GPT substep method considerably reduces the time discretization error. • No additional MC transport solutions are required within the time step. - Abstract: Coupled Monte Carlo (MC) methods are becoming widely used in reactor physics analysis and design. Many research groups therefore, developed their own coupled MC depletion codes. Typically, in such coupled code systems, neutron fluxes and cross sections are provided to the depletion module by solving a static neutron transport problem. These fluxes and cross sections are representative only of a specific time-point. In reality however, both quantities would change through the depletion time interval. Recently, Generalized Perturbation Theory (GPT) equivalent method that relies on collision history approach was implemented in Serpent MC code. This method was used here to calculate the sensitivity of each nuclide and reaction cross section due to the change in concentration of every isotope in the system. The coupling method proposed in this study also uses the substep approach, which incorporates these sensitivity coefficients to account for temporal changes in cross sections. As a result, a notable improvement in time dependent cross section behavior was obtained. The method was implemented in a wrapper script that couples Serpent with an external depletion solver. The performance of this method was compared with other existing methods. The results indicate that the proposed method requires substantially less MC transport solutions to achieve the same accuracy.

  3. Application of software quality assurance methods in validation and maintenance of reactor analysis computer codes

    International Nuclear Information System (INIS)

    Reznik, L.

    1994-01-01

    Various computer codes employed at Israel Electricity Company for preliminary reactor design analysis and fuel cycle scoping calculations have been often subject to program source modifications. Although most changes were due to computer or operating system compatibility problems, a number of significant modifications were due to model improvement and enhancements of algorithm efficiency and accuracy. With growing acceptance of software quality assurance requirements and methods, a program of implementing extensive testing of modified software has been adopted within the regular maintenance activities. In this work survey has been performed of various software quality assurance methods of software testing which belong mainly to the two major categories of implementation ('white box') and specification-based ('black box') testing. The results of this survey exhibits a clear preference of specification-based testing. In particular the equivalence class partitioning method and the boundary value method have been selected as especially suitable functional methods for testing reactor analysis codes.A separate study of software quality assurance methods and techniques has been performed in this work objective to establish appropriate pre-test software specification methods. Two methods of software analysis and specification have been selected as the most suitable for this purpose: The method of data flow diagrams has been shown to be particularly valuable for performing the functional/procedural software specification while the entities - relationship diagrams has been approved to be efficient for specifying software data/information domain. Feasibility of these two methods has been analyzed in particular for software uncertainty analysis and overall code accuracy estimation. (author). 14 refs

  4. The sensitivity analysis by adjoint method for the uncertainty evaluation of the CATHARE-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Barre, F.; de Crecy, A.; Perret, C. [French Atomic Energy Commission (CEA), Grenoble (France)

    1995-09-01

    This paper presents the application of the DASM (Discrete Adjoint Sensitivity Method) to CATHARE 2 thermal-hydraulics code. In a first part, the basis of this method is presented. The mathematical model of the CATHARE 2 code is based on the two fluid six equation model. It is discretized using implicit time discretization and it is relatively easy to implement this method in the code. The DASM is the ASM directly applied to the algebraic system of the discretized code equations which has been demonstrated to be the only solution of the mathematical model. The ASM is an integral part of the new version 1.4 of CATHARE. It acts as a post-processing module. It has been qualified by comparison with the {open_quotes}brute force{close_quotes} technique. In a second part, an application of the DASM in CATHARE 2 is presented. It deals with the determination of the uncertainties of the constitutive relationships, which is a compulsory step for calculating the final uncertainty of a given response. First, the general principles of the method are explained: the constitutive relationship are represented by several parameters and the aim is to calculate the variance-covariance matrix of these parameters. The experimental results of the separate effect tests used to establish the correlation are considered. The variance of the corresponding results calculated by CATHARE are estimated by comparing experiment and calculation. A DASM calculation is carried out to provide the derivatives of the responses. The final covariance matrix is obtained by combination of the variance of the responses and those derivatives. Then, the application of this method to a simple case-the blowdown Canon experiment-is presented. This application has been successfully performed.

  5. The sensitivity analysis by adjoint method for the uncertainty evaluation of the CATHARE-2 code

    International Nuclear Information System (INIS)

    Barre, F.; de Crecy, A.; Perret, C.

    1995-01-01

    This paper presents the application of the DASM (Discrete Adjoint Sensitivity Method) to CATHARE 2 thermal-hydraulics code. In a first part, the basis of this method is presented. The mathematical model of the CATHARE 2 code is based on the two fluid six equation model. It is discretized using implicit time discretization and it is relatively easy to implement this method in the code. The DASM is the ASM directly applied to the algebraic system of the discretized code equations which has been demonstrated to be the only solution of the mathematical model. The ASM is an integral part of the new version 1.4 of CATHARE. It acts as a post-processing module. It has been qualified by comparison with the open-quotes brute forceclose quotes technique. In a second part, an application of the DASM in CATHARE 2 is presented. It deals with the determination of the uncertainties of the constitutive relationships, which is a compulsory step for calculating the final uncertainty of a given response. First, the general principles of the method are explained: the constitutive relationship are represented by several parameters and the aim is to calculate the variance-covariance matrix of these parameters. The experimental results of the separate effect tests used to establish the correlation are considered. The variance of the corresponding results calculated by CATHARE are estimated by comparing experiment and calculation. A DASM calculation is carried out to provide the derivatives of the responses. The final covariance matrix is obtained by combination of the variance of the responses and those derivatives. Then, the application of this method to a simple case-the blowdown Canon experiment-is presented. This application has been successfully performed

  6. Benchmarking of epithermal methods in the lattice-physics code EPRI-CELL

    International Nuclear Information System (INIS)

    Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.; Toney, B.

    1982-01-01

    The epithermal cross section shielding methods used in the lattice physics code EPRI-CELL (E-C) have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology. These include: treatment of the external moderator source with intermediate resonance (IR) theory, development of a new Dancoff factor expression to account for clad interactions, development of a new method for treating resonance interference, and application of a generalized least squares method to compute best-estimate values for the Bell factor and group-dependent IR parameters. The modified E-C code with its new ENDF/B-V cross section library is tested for several numerical benchmark problems. Integral parameters computed by EC are compared with those obtained with point-cross section Monte Carlo calculations, and E-C fine group cross sections are benchmarked against point-cross section descrete ordinates calculations. It is found that the code modifications improve agreement between E-C and the more sophisticated methods. E-C shows excellent agreement on the integral parameters and usually agrees within a few percent on fine-group, shielded cross sections

  7. A novel quantum LSB-based steganography method using the Gray code for colored quantum images

    Science.gov (United States)

    Heidari, Shahrokh; Farzadnia, Ehsan

    2017-10-01

    As one of the prevalent data-hiding techniques, steganography is defined as the act of concealing secret information in a cover multimedia encompassing text, image, video and audio, imperceptibly, in order to perform interaction between the sender and the receiver in which nobody except the receiver can figure out the secret data. In this approach a quantum LSB-based steganography method utilizing the Gray code for quantum RGB images is investigated. This method uses the Gray code to accommodate two secret qubits in 3 LSBs of each pixel simultaneously according to reference tables. Experimental consequences which are analyzed in MATLAB environment, exhibit that the present schema shows good performance and also it is more secure and applicable than the previous one currently found in the literature.

  8. Performance of Multilevel Coding Schemes with Different Decoding Methods and Mapping Strategies in Mobile Fading Channels

    Institute of Scientific and Technical Information of China (English)

    YUAN Dongfeng; WANG Chengxiang; YAO Qi; CAO Zhigang

    2001-01-01

    Based on "capacity rule", the perfor-mance of multilevel coding (MLC) schemes with dif-ferent set partitioning strategies and decoding meth-ods in AWGN and Rayleigh fading channels is investi-gated, in which BCH codes are chosen as componentcodes and 8ASK modulation is used. Numerical re-sults indicate that MLC scheme with UP strategy canobtain optimal performance in AWGN channels andBP is the best mapping strategy for Rayleigh fadingchannels. BP strategy is of good robustness in bothkinds of channels to realize an optimum MLC system.Multistage decoding (MSD) is a sub-optimal decodingmethod of MLC for both channels. For Ungerboeckpartitioning (UP) and mixed partitioning (MP) strat-egy, MSD is strongly recommended to use for MLCsystem, while for BP strategy, PDL is suggested to useas a simple decoding method compared with MSD.

  9. Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.; Corella, M.R.; Esteban, A.; Martinez-Val, J.M.; Minguez, E.; Perlado, J.M.; Pena, J.; Matias, E. de; Llorente, A.; Navascues, J.; Serrano, J.

    1976-01-01

    Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author) [es

  10. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications

    International Nuclear Information System (INIS)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs

  11. How could the replica method improve accuracy of performance assessment of channel coding?

    Energy Technology Data Exchange (ETDEWEB)

    Kabashima, Yoshiyuki [Department of Computational Intelligence and Systems Science, Tokyo Institute of technology, Yokohama 226-8502 (Japan)], E-mail: kaba@dis.titech.ac.jp

    2009-12-01

    We explore the relation between the techniques of statistical mechanics and information theory for assessing the performance of channel coding. We base our study on a framework developed by Gallager in IEEE Trans. Inform. Theory IT-11, 3 (1965), where the minimum decoding error probability is upper-bounded by an average of a generalized Chernoff's bound over a code ensemble. We show that the resulting bound in the framework can be directly assessed by the replica method, which has been developed in statistical mechanics of disordered systems, whereas in Gallager's original methodology further replacement by another bound utilizing Jensen's inequality is necessary. Our approach associates a seemingly ad hoc restriction with respect to an adjustable parameter for optimizing the bound with a phase transition between two replica symmetric solutions, and can improve the accuracy of performance assessments of general code ensembles including low density parity check codes, although its mathematical justification is still open.

  12. Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  13. Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes

    CERN Document Server

    Mainardi, E; Donahue, R J

    2002-01-01

    The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using ...

  14. Transcoding method from H.264/AVC to high efficiency video coding based on similarity of intraprediction, interprediction, and motion vector

    Science.gov (United States)

    Liu, Mei-Feng; Zhong, Guo-Yun; He, Xiao-Hai; Qing, Lin-Bo

    2016-09-01

    Currently, most video resources on line are encoded in the H.264/AVC format. More fluent video transmission can be obtained if these resources are encoded in the newest international video coding standard: high efficiency video coding (HEVC). In order to improve the video transmission and storage on line, a transcoding method from H.264/AVC to HEVC is proposed. In this transcoding algorithm, the coding information of intraprediction, interprediction, and motion vector (MV) in H.264/AVC video stream are used to accelerate the coding in HEVC. It is found through experiments that the region of interprediction in HEVC overlaps that in H.264/AVC. Therefore, the intraprediction for the region in HEVC, which is interpredicted in H.264/AVC, can be skipped to reduce coding complexity. Several macroblocks in H.264/AVC are combined into one PU in HEVC when the MV difference between two of the macroblocks in H.264/AVC is lower than a threshold. This method selects only one coding unit depth and one prediction unit (PU) mode to reduce the coding complexity. An MV interpolation method of combined PU in HEVC is proposed according to the areas and distances between the center of one macroblock in H.264/AVC and that of the PU in HEVC. The predicted MV accelerates the motion estimation for HEVC coding. The simulation results show that our proposed algorithm achieves significant coding time reduction with a little loss in bitrates distortion rate, compared to the existing transcoding algorithms and normal HEVC coding.

  15. Simulation of clinical X-ray tube using the Monte Carlo Method - PENELOPE code

    International Nuclear Information System (INIS)

    Albuquerque, M.A.G.; David, M.G.; Almeida, C.E. de; Magalhaes, L.A.G.; Braz, D.

    2015-01-01

    Breast cancer is the most common type of cancer among women. The main strategy to increase the long-term survival of patients with this disease is the early detection of the tumor, and mammography is the most appropriate method for this purpose. Despite the reduction of cancer deaths, there is a big concern about the damage caused by the ionizing radiation to the breast tissue. To evaluate these measures it was modeled a mammography equipment, and obtained the depth spectra using the Monte Carlo method - PENELOPE code. The average energies of the spectra in depth and the half value layer of the mammography output spectrum. (author)

  16. Methods and codes for neutronic calculations of the MARIA research reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

    1998-01-01

    The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

  17. Methods for Using Small Non-Coding RNAs to Improve Recombinant Protein Expression in Mammalian Cells

    Directory of Open Access Journals (Sweden)

    Sarah Inwood

    2018-01-01

    Full Text Available The ability to produce recombinant proteins by utilizing different “cell factories” revolutionized the biotherapeutic and pharmaceutical industry. Chinese hamster ovary (CHO cells are the dominant industrial producer, especially for antibodies. Human embryonic kidney cells (HEK, while not being as widely used as CHO cells, are used where CHO cells are unable to meet the needs for expression, such as growth factors. Therefore, improving recombinant protein expression from mammalian cells is a priority, and continuing effort is being devoted to this topic. Non-coding RNAs are RNA segments that are not translated into a protein and often have a regulatory role. Since their discovery, major progress has been made towards understanding their functions. Non-coding RNA has been investigated extensively in relation to disease, especially cancer, and recently they have also been used as a method for engineering cells to improve their protein expression capability. In this review, we provide information about methods used to identify non-coding RNAs with the potential of improving recombinant protein expression in mammalian cell lines.

  18. Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

    International Nuclear Information System (INIS)

    Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.

    2014-01-01

    Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff

  19. Application Of WIMS Code To Calculation Kartini Reactor Parameters By Pin-Cell And Cluster Method

    International Nuclear Information System (INIS)

    Sumarsono, Bambang; Tjiptono, T.W.

    1996-01-01

    Analysis UZrH fuel element parameters calculation in Kartini Reactor by WIMS Code has been done. The analysis is done by pin cell and cluster method. The pin cell method is done as a function percent burn-up and by 8 group 3 region analysis and cluster method by 8 group 12 region analysis. From analysis and calculation resulted K ∼ = 1.3687 by pin cell method and K ∼ = 1.3162 by cluster method and so deviation is 3.83%. By pin cell analysis as a function percent burn-up at the percent burn-up greater than 59.50%, the multiplication factor is less than one (k ∼ < 1) it is mean that the fuel element reactivity is negative

  20. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1996-03-01

    During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

  1. A Mixed Methods Approach to Code Stakeholder Beliefs in Urban Water Governance

    Science.gov (United States)

    Bell, E. V.; Henry, A.; Pivo, G.

    2017-12-01

    What is a reliable way to code policies to represent belief systems? The Advocacy Coalition Framework posits that public policy may be viewed as manifestations of belief systems. Belief systems include both ontological beliefs about cause-and-effect relationships and policy effectiveness, as well as normative beliefs about appropriate policy instruments and the relative value of different outcomes. The idea that belief systems are embodied in public policy is important for urban water governance because it trains our focus on belief conflict; this can help us understand why many water-scarce cities do not adopt innovative technology despite available scientific information. To date, there has been very little research on systematic, rigorous methods to measure the belief system content of public policies. We address this by testing the relationship between beliefs and policy participation to develop an innovative coding framework. With a focus on urban water governance in Tucson, Arizona, we analyze grey literature on local water management. Mentioned policies are coded into a typology of common approaches identified in urban water governance literature, which include regulation, education, price and non-price incentives, green infrastructure and other types of technology. We then survey local water stakeholders about their perceptions of these policies. Urban water governance requires coordination of organizations from multiple sectors, and we cannot assume that belief development and policy participation occur in a vacuum. Thus, we use a generalized exponential random graph model to test the relationship between perceptions and policy participation in the Tucson water governance network. We measure policy perceptions for organizations by averaging across their respective, affiliated respondents and generating a belief distance matrix of coordinating network participants. Similarly, we generate a distance matrix of these actors based on the frequency of their

  2. An improved correlated sampling method for calculating correction factor of detector

    International Nuclear Information System (INIS)

    Wu Zhen; Li Junli; Cheng Jianping

    2006-01-01

    In the case of a small size detector lying inside a bulk of medium, there are two problems in the correction factors calculation of the detectors. One is that the detector is too small for the particles to arrive at and collide in; the other is that the ratio of two quantities is not accurate enough. The method discussed in this paper, which combines correlated sampling with modified particle collision auto-importance sampling, and has been realized on the MCNP-4C platform, can solve these two problems. Besides, other 3 variance reduction techniques are also combined with correlated sampling respectively to calculate a simple calculating model of the correction factors of detectors. The results prove that, although all the variance reduction techniques combined with correlated sampling can improve the calculating efficiency, the method combining the modified particle collision auto-importance sampling with the correlated sampling is the most efficient one. (authors)

  3. Non-linear heat transfer computer code by finite element method

    International Nuclear Information System (INIS)

    Nagato, Kotaro; Takikawa, Noboru

    1977-01-01

    The computer code THETA-2D for the calculation of temperature distribution by the two-dimensional finite element method was made for the analysis of heat transfer in a high temperature structure. Numerical experiment was performed for the numerical integration of the differential equation of heat conduction. The Runge-Kutta method of the numerical experiment produced an unstable solution. A stable solution was obtained by the β method with the β value of 0.35. In high temperature structures, the radiative heat transfer can not be neglected. To introduce a term of the radiative heat transfer, a functional neglecting the radiative heat transfer was derived at first. Then, the radiative term was added after the discretion by variation method. Five model calculations were carried out by the computer code. Calculation of steady heat conduction was performed. When estimated initial temperature is 1,000 degree C, reasonable heat blance was obtained. In case of steady-unsteady temperature calculation, the time integral by THETA-2D turned out to be under-estimation for enthalpy change. With a one-dimensional model, the temperature distribution in a structure, in which heat conductivity is dependent on temperature, was calculated. Calculation with a model which has a void inside was performed. Finally, model calculation for a complex system was carried out. (Kato, T.)

  4. Coding Model and Mapping Method of Spherical Diamond Discrete Grids Based on Icosahedron

    Directory of Open Access Journals (Sweden)

    LIN Bingxian

    2016-12-01

    Full Text Available Discrete Global Grid(DGG provides a fundamental environment for global-scale spatial data's organization and management. DGG's encoding scheme, which blocks coordinate transformation between different coordination reference frames and reduces the complexity of spatial analysis, contributes a lot to the multi-scale expression and unified modeling of spatial data. Compared with other kinds of DGGs, Diamond Discrete Global Grid(DDGG based on icosahedron is beneficial to the spherical spatial data's integration and expression for much better geometric properties. However, its structure seems more complicated than DDGG on octahedron due to its initial diamond's edges cannot fit meridian and parallel. New challenges are posed when it comes to the construction of hierarchical encoding system and mapping relationship with geographic coordinates. On this issue, this paper presents a DDGG's coding system based on the Hilbert curve and designs conversion methods between codes and geographical coordinates. The study results indicate that this encoding system based on the Hilbert curve can express space scale and location information implicitly with the similarity between DDG and planar grid put into practice, and balances efficiency and accuracy of conversion between codes and geographical coordinates in order to support global massive spatial data's modeling, integrated management and all kinds of spatial analysis.

  5. An asynchronous writing method for restart files in the gysela code in prevision of exascale systems*

    Directory of Open Access Journals (Sweden)

    Thomine O.

    2013-12-01

    Full Text Available The present work deals with an optimization procedure developed in the full-f global GYrokinetic SEmi-LAgrangian code (GYSELA. Optimizing the writing of the restart files is necessary to reduce the computing impact of crashes. These files require a very large memory space, and particularly so for very large mesh sizes. The limited bandwidth of the data pipe between the comput- ing nodes and the storage system induces a non-scalable part in the GYSELA code, which increases with the mesh size. Indeed the transfer time of RAM to data depends linearly on the files size. The necessity of non synchronized writing-in-file procedure is therefore crucial. A new GYSELA module has been developed. This asynchronous procedure allows the frequent writ- ing of the restart files, whilst preventing a severe slowing down due to the limited writing bandwidth. This method has been improved to generate a checksum control of the restart files, and automatically rerun the code in case of a crash for any cause.

  6. Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)

  7. Implementation of the probability table method in a continuous-energy Monte Carlo code system

    International Nuclear Information System (INIS)

    Sutton, T.M.; Brown, F.B.

    1998-10-01

    RACER is a particle-transport Monte Carlo code that utilizes a continuous-energy treatment for neutrons and neutron cross section data. Until recently, neutron cross sections in the unresolved resonance range (URR) have been treated in RACER using smooth, dilute-average representations. This paper describes how RACER has been modified to use probability tables to treat cross sections in the URR, and the computer codes that have been developed to compute the tables from the unresolved resonance parameters contained in ENDF/B data files. A companion paper presents results of Monte Carlo calculations that demonstrate the effect of the use of probability tables versus the use of dilute-average cross sections for the URR. The next section provides a brief review of the probability table method as implemented in the RACER system. The production of the probability tables for use by RACER takes place in two steps. The first step is the generation of probability tables from the nuclear parameters contained in the ENDF/B data files. This step, and the code written to perform it, are described in Section 3. The tables produced are at energy points determined by the ENDF/B parameters and/or accuracy considerations. The tables actually used in the RACER calculations are obtained in the second step from those produced in the first. These tables are generated at energy points specific to the RACER calculation. Section 4 describes this step and the code written to implement it, as well as modifications made to RACER to enable it to use the tables. Finally, some results and conclusions are presented in Section 5

  8. CHF predictor derived from a 3D thermal-hydraulic code and an advanced statistical method

    International Nuclear Information System (INIS)

    Banner, D.; Aubry, S.

    2004-01-01

    A rod bundle CHF predictor has been determined by using a 3D code (THYC) to compute local thermal-hydraulic conditions at the boiling crisis location. These local parameters have been correlated to the critical heat flux by using an advanced statistical method based on spline functions. The main characteristics of the predictor are presented in conjunction with a detailed analysis of predictions (P/M ratio) in order to prove that the usual safety methodology can be applied with such a predictor. A thermal-hydraulic design criterion is obtained (1.13) and the predictor is compared with the WRB-1 correlation. (author)

  9. Inverse Heat Conduction Methods in the CHAR Code for Aerothermal Flight Data Reconstruction

    Science.gov (United States)

    Oliver, A. Brandon; Amar, Adam J.

    2016-01-01

    Reconstruction of flight aerothermal environments often requires the solution of an inverse heat transfer problem, which is an ill-posed problem of determining boundary conditions from discrete measurements in the interior of the domain. This paper will present the algorithms implemented in the CHAR code for use in reconstruction of EFT-1 flight data and future testing activities. Implementation details will be discussed, and alternative hybrid-methods that are permitted by the implementation will be described. Results will be presented for a number of problems.

  10. Comparative evaluation of various optimization methods and the development of an optimization code system SCOOP

    International Nuclear Information System (INIS)

    Suzuki, Tadakazu

    1979-11-01

    Thirty two programs for linear and nonlinear optimization problems with or without constraints have been developed or incorporated, and their stability, convergence and efficiency have been examined. On the basis of these evaluations, the first version of the optimization code system SCOOP-I has been completed. The SCOOP-I is designed to be an efficient, reliable, useful and also flexible system for general applications. The system enables one to find global optimization point for a wide class of problems by selecting the most appropriate optimization method built in it. (author)

  11. A method of non-contact reading code based on computer vision

    Science.gov (United States)

    Zhang, Chunsen; Zong, Xiaoyu; Guo, Bingxuan

    2018-03-01

    With the purpose of guarantee the computer information exchange security between internal and external network (trusted network and un-trusted network), A non-contact Reading code method based on machine vision has been proposed. Which is different from the existing network physical isolation method. By using the computer monitors, camera and other equipment. Deal with the information which will be on exchanged, Include image coding ,Generate the standard image , Display and get the actual image , Calculate homography matrix, Image distort correction and decoding in calibration, To achieve the computer information security, Non-contact, One-way transmission between the internal and external network , The effectiveness of the proposed method is verified by experiments on real computer text data, The speed of data transfer can be achieved 24kb/s. The experiment shows that this algorithm has the characteristics of high security, fast velocity and less loss of information. Which can meet the daily needs of the confidentiality department to update the data effectively and reliably, Solved the difficulty of computer information exchange between Secret network and non-secret network, With distinctive originality, practicability, and practical research value.

  12. Review of solution approach, methods, and recent results of the TRAC-PF1 system code

    International Nuclear Information System (INIS)

    Mahaffy, J.H.; Liles, D.R.; Knight, T.D.

    1983-01-01

    The current version of the Transient Reactor Analysis Code (TRAC-PF1) was created to improve on the capabilities of its predecessor (TRAC-PD2) for analyzing slow reactor transients such as small-break loss-of-coolant accidents. TRAC-PF1 continues to use a semi-implicit finite-difference method for modeling three-dimensional flows in the reactor vessel. However, it contains a new stability-enhancing two-step (SETS) finite-difference tecnique for one-dimensional flow calculations. This method is not restricted by a material Courant stability condition, allowing much larger time-step sizes during slow transients than would a semi-implicit method. These have been successfully applied to the analysis of a variety of experiments and hypothetical plant transients covering a full range of two-phase flow regimes

  13. An imaging method of wavefront coding system based on phase plate rotation

    Science.gov (United States)

    Yi, Rigui; Chen, Xi; Dong, Liquan; Liu, Ming; Zhao, Yuejin; Liu, Xiaohua

    2018-01-01

    Wave-front coding has a great prospect in extending the depth of the optical imaging system and reducing optical aberrations, but the image quality and noise performance are inevitably reduced. According to the theoretical analysis of the wave-front coding system and the phase function expression of the cubic phase plate, this paper analyzed and utilized the feature that the phase function expression would be invariant in the new coordinate system when the phase plate rotates at different angles around the z-axis, and we proposed a method based on the rotation of the phase plate and image fusion. First, let the phase plate rotated at a certain angle around the z-axis, the shape and distribution of the PSF obtained on the image surface remain unchanged, the rotation angle and direction are consistent with the rotation angle of the phase plate. Then, the middle blurred image is filtered by the point spread function of the rotation adjustment. Finally, the reconstruction images were fused by the method of the Laplacian pyramid image fusion and the Fourier transform spectrum fusion method, and the results were evaluated subjectively and objectively. In this paper, we used Matlab to simulate the images. By using the Laplacian pyramid image fusion method, the signal-to-noise ratio of the image is increased by 19% 27%, the clarity is increased by 11% 15% , and the average gradient is increased by 4% 9% . By using the Fourier transform spectrum fusion method, the signal-to-noise ratio of the image is increased by 14% 23%, the clarity is increased by 6% 11% , and the average gradient is improved by 2% 6%. The experimental results show that the image processing by the above method can improve the quality of the restored image, improving the image clarity, and can effectively preserve the image information.

  14. Non-coding RNA detection methods combined to improve usability, reproducibility and precision

    Directory of Open Access Journals (Sweden)

    Kreikemeyer Bernd

    2010-09-01

    Full Text Available Abstract Background Non-coding RNAs gain more attention as their diverse roles in many cellular processes are discovered. At the same time, the need for efficient computational prediction of ncRNAs increases with the pace of sequencing technology. Existing tools are based on various approaches and techniques, but none of them provides a reliable ncRNA detector yet. Consequently, a natural approach is to combine existing tools. Due to a lack of standard input and output formats combination and comparison of existing tools is difficult. Also, for genomic scans they often need to be incorporated in detection workflows using custom scripts, which decreases transparency and reproducibility. Results We developed a Java-based framework to integrate existing tools and methods for ncRNA detection. This framework enables users to construct transparent detection workflows and to combine and compare different methods efficiently. We demonstrate the effectiveness of combining detection methods in case studies with the small genomes of Escherichia coli, Listeria monocytogenes and Streptococcus pyogenes. With the combined method, we gained 10% to 20% precision for sensitivities from 30% to 80%. Further, we investigated Streptococcus pyogenes for novel ncRNAs. Using multiple methods--integrated by our framework--we determined four highly probable candidates. We verified all four candidates experimentally using RT-PCR. Conclusions We have created an extensible framework for practical, transparent and reproducible combination and comparison of ncRNA detection methods. We have proven the effectiveness of this approach in tests and by guiding experiments to find new ncRNAs. The software is freely available under the GNU General Public License (GPL, version 3 at http://www.sbi.uni-rostock.de/moses along with source code, screen shots, examples and tutorial material.

  15. DEEP code to calculate dose equivalents in human phantom for external photon exposure by Monte Carlo method

    International Nuclear Information System (INIS)

    Yamaguchi, Yasuhiro

    1991-01-01

    The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)

  16. Construction method of QC-LDPC codes based on multiplicative group of finite field in optical communication

    Science.gov (United States)

    Huang, Sheng; Ao, Xiang; Li, Yuan-yuan; Zhang, Rui

    2016-09-01

    In order to meet the needs of high-speed development of optical communication system, a construction method of quasi-cyclic low-density parity-check (QC-LDPC) codes based on multiplicative group of finite field is proposed. The Tanner graph of parity check matrix of the code constructed by this method has no cycle of length 4, and it can make sure that the obtained code can get a good distance property. Simulation results show that when the bit error rate ( BER) is 10-6, in the same simulation environment, the net coding gain ( NCG) of the proposed QC-LDPC(3 780, 3 540) code with the code rate of 93.7% in this paper is improved by 2.18 dB and 1.6 dB respectively compared with those of the RS(255, 239) code in ITU-T G.975 and the LDPC(3 2640, 3 0592) code in ITU-T G.975.1. In addition, the NCG of the proposed QC-LDPC(3 780, 3 540) code is respectively 0.2 dB and 0.4 dB higher compared with those of the SG-QC-LDPC(3 780, 3 540) code based on the two different subgroups in finite field and the AS-QC-LDPC(3 780, 3 540) code based on the two arbitrary sets of a finite field. Thus, the proposed QC-LDPC(3 780, 3 540) code in this paper can be well applied in optical communication systems.

  17. An Effective Transform Unit Size Decision Method for High Efficiency Video Coding

    Directory of Open Access Journals (Sweden)

    Chou-Chen Wang

    2014-01-01

    Full Text Available High efficiency video coding (HEVC is the latest video coding standard. HEVC can achieve higher compression performance than previous standards, such as MPEG-4, H.263, and H.264/AVC. However, HEVC requires enormous computational complexity in encoding process due to quadtree structure. In order to reduce the computational burden of HEVC encoder, an early transform unit (TU decision algorithm (ETDA is adopted to pruning the residual quadtree (RQT at early stage based on the number of nonzero DCT coefficients (called NNZ-EDTA to accelerate the encoding process. However, the NNZ-ETDA cannot effectively reduce the computational load for sequences with active motion or rich texture. Therefore, in order to further improve the performance of NNZ-ETDA, we propose an adaptive RQT-depth decision for NNZ-ETDA (called ARD-NNZ-ETDA by exploiting the characteristics of high temporal-spatial correlation that exist in nature video sequences. Simulation results show that the proposed method can achieve time improving ratio (TIR about 61.26%~81.48% when compared to the HEVC test model 8.1 (HM 8.1 with insignificant loss of image quality. Compared with the NNZ-ETDA, the proposed method can further achieve an average TIR about 8.29%~17.92%.

  18. KIN SP: A boundary element method based code for single pile kinematic bending in layered soil

    Directory of Open Access Journals (Sweden)

    Stefano Stacul

    2018-02-01

    Full Text Available In high seismicity areas, it is important to consider kinematic effects to properly design pile foundations. Kinematic effects are due to the interaction between pile and soil deformations induced by seismic waves. One of the effect is the arise of significant strains in weak soils that induce bending moments on piles. These moments can be significant in presence of a high stiffness contrast in a soil deposit. The single pile kinematic interaction problem is generally solved with beam on dynamic Winkler foundation approaches (BDWF or using continuous models. In this work, a new boundary element method (BEM based computer code (KIN SP is presented where the kinematic analysis is preceded by a free-field response analysis. The analysis results of this method, in terms of bending moments at the pile-head and at the interface of a two-layered soil, are influenced by many factors including the soil–pile interface discretization. A parametric study is presented with the aim to suggest the minimum number of boundary elements to guarantee the accuracy of a BEM solution, for typical pile–soil relative stiffness values as a function of the pile diameter, the location of the interface of a two-layered soil and of the stiffness contrast. KIN SP results have been compared with simplified solutions in literature and with those obtained using a quasi-three-dimensional (3D finite element code.

  19. An Improved Real-Coded Population-Based Extremal Optimization Method for Continuous Unconstrained Optimization Problems

    Directory of Open Access Journals (Sweden)

    Guo-Qiang Zeng

    2014-01-01

    Full Text Available As a novel evolutionary optimization method, extremal optimization (EO has been successfully applied to a variety of combinatorial optimization problems. However, the applications of EO in continuous optimization problems are relatively rare. This paper proposes an improved real-coded population-based EO method (IRPEO for continuous unconstrained optimization problems. The key operations of IRPEO include generation of real-coded random initial population, evaluation of individual and population fitness, selection of bad elements according to power-law probability distribution, generation of new population based on uniform random mutation, and updating the population by accepting the new population unconditionally. The experimental results on 10 benchmark test functions with the dimension N=30 have shown that IRPEO is competitive or even better than the recently reported various genetic algorithm (GA versions with different mutation operations in terms of simplicity, effectiveness, and efficiency. Furthermore, the superiority of IRPEO to other evolutionary algorithms such as original population-based EO, particle swarm optimization (PSO, and the hybrid PSO-EO is also demonstrated by the experimental results on some benchmark functions.

  20. Methods tuned on the physical problem. A way to improve numerical codes

    International Nuclear Information System (INIS)

    Ixaru, L.Gr.

    2010-01-01

    We consider the problem on how the numerical methods tuned on the physical problem can contribute to the enhancement of the performance of the codes. We illustrate this on two simple cases: solution of time independent one-dimensional Schroedinger equation, and the computation of integrals with oscillatory integrands. In both cases the tuned versions bring a massive gain in accuracy at negligible extra cost. We presented two simple problems where successive levels of tuning enhance significantly the accuracy at negligible extra cost. These problems should be seen as representing only some illustrations on how the codes can be improved but we must also mention that in many cases tuned versions still have to be developed. Just for a suggestion, quadrature formulae which involve the integrand and a number of successive derivatives of this exist, but no formula is available when some of these derivatives are missing, for example when we dispose of y and y'' but not of y'. A direct application will be on the case when the integrand involves the solution of the Schrodinger equation by the method of Numerov. (author)

  1. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  2. SWAAM-LT: The long-term, sodium/water reaction analysis method computer code

    International Nuclear Information System (INIS)

    Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.

    1993-01-01

    The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data

  3. Comparison of a semi-empirical method with some model codes for gamma-ray spectrum calculation

    Energy Technology Data Exchange (ETDEWEB)

    Sheng, Fan; Zhixiang, Zhao [Chinese Nuclear Data Center, Beijing, BJ (China)

    1996-06-01

    Gamma-ray spectra calculated by a semi-empirical method are compared with those calculated by the model codes such as GNASH, TNG, UNF and NDCP-1. The results of the calculations are discussed. (2 tabs., 3 figs.).

  4. Stego Keys Performance on Feature Based Coding Method in Text Domain

    Directory of Open Access Journals (Sweden)

    Din Roshidi

    2017-01-01

    Full Text Available A main critical factor on embedding process in any text steganography method is a key used known as stego key. This factor will be influenced the success of the embedding process of text steganography method to hide a message from third party or any adversary. One of the important aspects on embedding process in text steganography method is the fitness performance of the stego key. Three parameters of the fitness performance of the stego key have been identified such as capacity ratio, embedded fitness ratio and saving space ratio. It is because a better as capacity ratio, embedded fitness ratio and saving space ratio offers of any stego key; a more message can be hidden. Therefore, main objective of this paper is to analyze three features coding based namely CALP, VERT and QUAD of stego keys in text steganography on their capacity ratio, embedded fitness ratio and saving space ratio. It is found that CALP method give a good effort performance compared to VERT and QUAD methods.

  5. The FLUKA code for application of Monte Carlo methods to promote high precision ion beam therapy

    CERN Document Server

    Parodi, K; Cerutti, F; Ferrari, A; Mairani, A; Paganetti, H; Sommerer, F

    2010-01-01

    Monte Carlo (MC) methods are increasingly being utilized to support several aspects of commissioning and clinical operation of ion beam therapy facilities. In this contribution two emerging areas of MC applications are outlined. The value of MC modeling to promote accurate treatment planning is addressed via examples of application of the FLUKA code to proton and carbon ion therapy at the Heidelberg Ion Beam Therapy Center in Heidelberg, Germany, and at the Proton Therapy Center of Massachusetts General Hospital (MGH) Boston, USA. These include generation of basic data for input into the treatment planning system (TPS) and validation of the TPS analytical pencil-beam dose computations. Moreover, we review the implementation of PET/CT (Positron-Emission-Tomography / Computed- Tomography) imaging for in-vivo verification of proton therapy at MGH. Here, MC is used to calculate irradiation-induced positron-emitter production in tissue for comparison with the +-activity measurement in order to infer indirect infor...

  6. Quantum image pseudocolor coding based on the density-stratified method

    Science.gov (United States)

    Jiang, Nan; Wu, Wenya; Wang, Luo; Zhao, Na

    2015-05-01

    Pseudocolor processing is a branch of image enhancement. It dyes grayscale images to color images to make the images more beautiful or to highlight some parts on the images. This paper proposes a quantum image pseudocolor coding scheme based on the density-stratified method which defines a colormap and changes the density value from gray to color parallel according to the colormap. Firstly, two data structures: quantum image GQIR and quantum colormap QCR are reviewed or proposed. Then, the quantum density-stratified algorithm is presented. Based on them, the quantum realization in the form of circuits is given. The main advantages of the quantum version for pseudocolor processing over the classical approach are that it needs less memory and can speed up the computation. Two kinds of examples help us to describe the scheme further. Finally, the future work are analyzed.

  7. ORIGEN-2.2, Isotope Generation and Depletion Code Matrix Exponential Method

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of problem or function: ORIGEN is a computer code system for calculating the buildup, decay, and processing of radioactive materials. ORIGEN2 is a revised version of ORIGEN and incorporates updates of the reactor models, cross sections, fission product yields, decay data, and decay photon data, as well as the source code. ORIGEN-2.1 replaces ORIGEN and includes additional libraries for standard and extended-burnup PWR and BWR calculations, which are documented in ORNL/TM-11018. ORIGEN2.1 was first released in August 1991 and was replaced with ORIGEN2 Version 2.2 in June 2002. Version 2.2 was the first update to ORIGEN2 in over 10 years and was stimulated by a user discovering a discrepancy in the mass of fission products calculated using ORIGEN2 V2.1. Code modifications, as well as reducing the irradiation time step to no more than 100 days/step reduced the discrepancy from ∼10% to 0.16%. The bug does not noticeably affect the fission product mass in typical ORIGEN2 calculations involving reactor fuels because essentially all of the fissions come from actinides that have explicit fission product yield libraries. Thus, most previous ORIGEN2 calculations that were otherwise set up properly should not be affected. 2 - Method of solution: ORIGEN uses a matrix exponential method to solve a large system of coupled, linear, first-order ordinary differential equations with constant coefficients. ORIGEN2 has been variably dimensioned to allow the user to tailor the size of the executable module to the problem size and/or the available computer space. Dimensioned arrays have been set large enough to handle almost any size problem, using virtual memory capabilities available on most mainframe and 386/486 based PCS. The user is provided with much of the framework necessary to put some of the arrays to several different uses, call for the subroutines that perform the desired operations, and provide a mechanism to execute multiple ORIGEN2 problems with a single

  8. A five-colour colour-coded mapping method for DCE-MRI analysis of head and neck tumours

    International Nuclear Information System (INIS)

    Yuan, J.; Chow, S.K.K.; Yeung, D.K.W.; King, A.D.

    2012-01-01

    Aim: To devise a method to convert the time–intensity curves (TICs) of head and neck dynamic contrast-enhanced (DCE) magnetic resonance imaging (MRI) data into a pixel-by-pixel colour-coded map for identifying normal tissues and tumours. Materials and methods: Twenty-three patients with head and neck squamous cell carcinoma (HNSCC) underwent DCE-MRI. TIC patterns of primary tumours, metastatic nodes, and normal tissues were assessed and a program was devised to convert the patterns into a classified colour-coded map. The enhancement patterns of tumours and normal tissue structures were evaluated and categorized into nine grades (0–8) based on the predominance of coloured pixels on maps. Results: Five identified TIC patterns were converted into a colour-coded map consisting of red (maximum enhancement), brown (continuous slow rise-up), yellow (rapid wash-in and wash-out), green (rapid wash-in and plateau), and blue (rapid wash-in and rise-up). The colour-coded map distinguished all 21 primary tumours and 15 metastatic nodes from normal structures. Primary tumours and metastatic nodes were colour coded as predominantly yellow (grades 1–2) in 17/21 and 6/15, green (grades 3–5) in 3/21 and 5/15, and blue (grades 6–7) in 1/21 and 4/15, respectively. Vessels were coded red in 46/46 (grade 0) and muscles were coded brown in 23/23 (grade 8). Salivary glands, thyroid glands, and palatine tonsils were coded into predominantly yellow (grade 1) in 46/46 and 10/10 and 18/22, respectively. Conclusion: DCE-MRI derived five-colour-coded mapping provides an objective easy-to-interpret method to assess the dynamic enhancement pattern of head and neck cancers.

  9. Embedded 3D shape measurement system based on a novel spatio-temporal coding method

    Science.gov (United States)

    Xu, Bin; Tian, Jindong; Tian, Yong; Li, Dong

    2016-11-01

    Structured light measurement has been wildly used since 1970s in industrial component detection, reverse engineering, 3D molding, robot navigation, medical and many other fields. In order to satisfy the demand for high speed, high precision and high resolution 3-D measurement for embedded system, a new patterns combining binary and gray coding principle in space are designed and projected onto the object surface orderly. Each pixel corresponds to the designed sequence of gray values in time - domain, which is treated as a feature vector. The unique gray vector is then dimensionally reduced to a scalar which could be used as characteristic information for binocular matching. In this method, the number of projected structured light patterns is reduced, and the time-consuming phase unwrapping in traditional phase shift methods is avoided. This algorithm is eventually implemented on DM3730 embedded system for 3-D measuring, which consists of an ARM and a DSP core and has a strong capability of digital signal processing. Experimental results demonstrated the feasibility of the proposed method.

  10. Development of an aeroelastic code based on three-dimensional viscous–inviscid method for wind turbine computations

    DEFF Research Database (Denmark)

    Sessarego, Matias; Ramos García, Néstor; Sørensen, Jens Nørkær

    2017-01-01

    Aerodynamic and structural dynamic performance analysis of modern wind turbines are routinely estimated in the wind energy field using computational tools known as aeroelastic codes. Most aeroelastic codes use the blade element momentum (BEM) technique to model the rotor aerodynamics and a modal......, multi-body or the finite-element approach to model the turbine structural dynamics. The present work describes the development of a novel aeroelastic code that combines a three-dimensional viscous–inviscid interactive method, method for interactive rotor aerodynamic simulations (MIRAS...... Code Comparison Collaboration Project. Simulation tests consist of steady wind inflow conditions with different combinations of yaw error, wind shear, tower shadow and turbine-elastic modeling. Turbulent inflow created by using a Mann box is also considered. MIRAS-FLEX results, such as blade tip...

  11. Analysis of mixed oxides critical experiments using the Hammer-Technion code with self-shielding treatment by Bondarenko method

    International Nuclear Information System (INIS)

    Abe, Alfredo Y.; Santos, Adimir dos

    1995-01-01

    The present work summarizes the verification of the treatment of self-shielding based on Bondarenko method in HAMMER-TECHNION cell code for the Pu O 2 -U O 2 critical system using JENDL-3 nuclear data library. The results obtained are in excellent agreement with the original treatment of self-shielding employed by HAMMER-TECHNION cell code. (author). 9 refs, 1 fig, 9 tabs

  12. Utilisation of best estimate system codes and best estimate methods in safety analyses of VVER reactors in the Czech Republic

    International Nuclear Information System (INIS)

    Macek, Jiri; Kral, Pavel

    2010-01-01

    The content of the presentation was as follows: Conservative versus best estimate approach, Brief description and selection of methodology, Description of uncertainty methods, Examples of the BE methodology. It is concluded that where BE computer codes are used, uncertainty and sensitivity analyses should be included; if best estimate codes + uncertainty are used, the safety margins increase; and BE + BSA is the next step in licensing analyses. (P.A.)

  13. Criticality Analysis Of TCA Critical Lattices With MNCP-4C Monte Carlo Calculation

    International Nuclear Information System (INIS)

    Zuhair

    2002-01-01

    The use of uranium-plutonium mixed oxide (MOX) fuel in electric generation light water reactor (PWR, BWR) is being planned in Japan. Therefore, the accuracy evaluations of neutronic analysis code for MOX cores have been employed by many scientists and reactor physicists. Benchmark evaluations for TCA was done using various calculation methods. The Monte Carlo become the most reliable method to predict criticality of various reactor types. In this analysis, the MCNP-4C code was chosen because various superiorities the code has. All in all, the MCNP-4C calculation for TCA core with 38 MOX critical lattice configurations gave the results with high accuracy. The JENDL-3.2 library showed significantly closer results to the ENDF/B-V. The k eff values calculated with the ENDF/B-VI library gave underestimated results. The ENDF/B-V library gave the best estimation. It can be concluded that MCNP-4C calculation, especially with ENDF/B-V and JENDL-3.2 libraries, for MOX fuel utilized NPP design in reactor core is the best choice

  14. Iterative linear solvers in a 2D radiation-hydrodynamics code: Methods and performance

    International Nuclear Information System (INIS)

    Baldwin, C.; Brown, P.N.; Falgout, R.; Graziani, F.; Jones, J.

    1999-01-01

    Computer codes containing both hydrodynamics and radiation play a central role in simulating both astrophysical and inertial confinement fusion (ICF) phenomena. A crucial aspect of these codes is that they require an implicit solution of the radiation diffusion equations. The authors present in this paper the results of a comparison of five different linear solvers on a range of complex radiation and radiation-hydrodynamics problems. The linear solvers used are diagonally scaled conjugate gradient, GMRES with incomplete LU preconditioning, conjugate gradient with incomplete Cholesky preconditioning, multigrid, and multigrid-preconditioned conjugate gradient. These problems involve shock propagation, opacities varying over 5--6 orders of magnitude, tabular equations of state, and dynamic ALE (Arbitrary Lagrangian Eulerian) meshes. They perform a problem size scalability study by comparing linear solver performance over a wide range of problem sizes from 1,000 to 100,000 zones. The fundamental question they address in this paper is: Is it more efficient to invert the matrix in many inexpensive steps (like diagonally scaled conjugate gradient) or in fewer expensive steps (like multigrid)? In addition, what is the answer to this question as a function of problem size and is the answer problem dependent? They find that the diagonally scaled conjugate gradient method performs poorly with the growth of problem size, increasing in both iteration count and overall CPU time with the size of the problem and also increasing for larger time steps. For all problems considered, the multigrid algorithms scale almost perfectly (i.e., the iteration count is approximately independent of problem size and problem time step). For pure radiation flow problems (i.e., no hydrodynamics), they see speedups in CPU time of factors of ∼15--30 for the largest problems, when comparing the multigrid solvers relative to diagonal scaled conjugate gradient

  15. Monte-Carlo method - codes for the study of criticality problems (on IBM 7094); Methode de Monte- Carlo - codes pour l'etude des problemes de criticite (IBM 7094)

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, J; Rabot, H; Robin, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The two codes presented here allow to determine the multiplication constant of media containing fissionable materials under numerous and divided forms; they are based on the Monte-Carlo method. The first code apply to x, y, z, geometries. The volume to be studied ought to be divisible in parallelepipeds, the media within each parallelepiped being limited by non-secant surfaces. The second code is intended for r, 0, z geometries. The results include an analysis of collisions in each medium. Applications and examples with informations on time and accuracy are given. (authors) [French] Les deux codes presentes dans ce rapport permettent la determination des coefficients de multiplication de milieux contenant des matieres fissiles sous des formes tres variees et divisees, ils reposent sur la methode de Monte-Carlo. Le premier code s'applique aux geometries x, y, z, le volume a etudier doit pouvoir etre decompose en parallelepipedes, les milieux a l'interieur de chaque parallelepipede etant limites par des surfaces non secantes. Le deuxieme code s'applique aux geometries r, 0, z. Les resultats comportent une analyse des collisions dans chaque milieu. Des applications et des exemples avec les indications de temps et de precision sont fournis. (auteurs)

  16. Green's function method and its application to verification of diffusion models of GASFLOW code

    International Nuclear Information System (INIS)

    Xu, Z.; Travis, J.R.; Breitung, W.

    2007-07-01

    To validate the diffusion model and the aerosol particle model of the GASFLOW computer code, theoretical solutions of advection diffusion problems are developed by using the Green's function method. The work consists of a theory part and an application part. In the first part, the Green's functions of one-dimensional advection diffusion problems are solved in infinite, semi-infinite and finite domains with the Dirichlet, the Neumann and/or the Robin boundary conditions. Novel and effective image systems especially for the advection diffusion problems are made to find the Green's functions in a semi-infinite domain. Eigenfunction method is utilized to find the Green's functions in a bounded domain. In the case, key steps of a coordinate transform based on a concept of reversed time scale, a Laplace transform and an exponential transform are proposed to solve the Green's functions. Then the product rule of the multi-dimensional Green's functions is discussed in a Cartesian coordinate system. Based on the building blocks of one-dimensional Green's functions, the multi-dimensional Green's function solution can be constructed by applying the product rule. Green's function tables are summarized to facilitate the application of the Green's function. In the second part, the obtained Green's function solutions benchmark a series of validations to the diffusion model of gas species in continuous phase and the diffusion model of discrete aerosol particles in the GASFLOW code. Perfect agreements are obtained between the GASFLOW simulations and the Green's function solutions in case of the gas diffusion. Very good consistencies are found between the theoretical solutions of the advection diffusion equations and the numerical particle distributions in advective flows, when the drag force between the micron-sized particles and the conveying gas flow meets the Stokes' law about resistance. This situation is corresponding to a very small Reynolds number based on the particle

  17. Decoy state method for quantum cryptography based on phase coding into faint laser pulses

    Science.gov (United States)

    Kulik, S. P.; Molotkov, S. N.

    2017-12-01

    We discuss the photon number splitting attack (PNS) in systems of quantum cryptography with phase coding. It is shown that this attack, as well as the structural equations for the PNS attack for phase encoding, differs physically from the analogous attack applied to the polarization coding. As far as we know, in practice, in all works to date processing of experimental data has been done for phase coding, but using formulas for polarization coding. This can lead to inadequate results for the length of the secret key. These calculations are important for the correct interpretation of the results, especially if it concerns the criterion of secrecy in quantum cryptography.

  18. Novel methods in the Particle-In-Cell accelerator Code-Framework Warp

    Energy Technology Data Exchange (ETDEWEB)

    Vay, J-L [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Grote, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Cohen, R. H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Friedman, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2012-12-26

    The Particle-In-Cell (PIC) Code-Framework Warp is being developed by the Heavy Ion Fusion Science Virtual National Laboratory (HIFS-VNL) to guide the development of accelerators that can deliver beams suitable for high-energy density experiments and implosion of inertial fusion capsules. It is also applied in various areas outside the Heavy Ion Fusion program to the study and design of existing and next-generation high-energy accelerators, including the study of electron cloud effects and laser wakefield acceleration for example. This study presents an overview of Warp's capabilities, summarizing recent original numerical methods that were developed by the HIFS-VNL (including PIC with adaptive mesh refinement, a large-timestep 'drift-Lorentz' mover for arbitrarily magnetized species, a relativistic Lorentz invariant leapfrog particle pusher, simulations in Lorentz-boosted frames, an electromagnetic solver with tunable numerical dispersion and efficient stride-based digital filtering), with special emphasis on the description of the mesh refinement capability. In addition, selected examples of the applications of the methods to the abovementioned fields are given.

  19. RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes

    International Nuclear Information System (INIS)

    Parisi, C.; D'Auria, F.

    2007-01-01

    The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)

  20. Uncertainty analysis methods for quantification of source terms using a large computer code

    International Nuclear Information System (INIS)

    Han, Seok Jung

    1997-02-01

    Quantification of uncertainties in the source term estimations by a large computer code, such as MELCOR and MAAP, is an essential process of the current probabilistic safety assessments (PSAs). The main objectives of the present study are (1) to investigate the applicability of a combined procedure of the response surface method (RSM) based on input determined from a statistical design and the Latin hypercube sampling (LHS) technique for the uncertainty analysis of CsI release fractions under a hypothetical severe accident sequence of a station blackout at Young-Gwang nuclear power plant using MAAP3.0B code as a benchmark problem; and (2) to propose a new measure of uncertainty importance based on the distributional sensitivity analysis. On the basis of the results obtained in the present work, the RSM is recommended to be used as a principal tool for an overall uncertainty analysis in source term quantifications, while using the LHS in the calculations of standardized regression coefficients (SRC) and standardized rank regression coefficients (SRRC) to determine the subset of the most important input parameters in the final screening step and to check the cumulative distribution functions (cdfs) obtained by RSM. Verification of the response surface model for its sufficient accuracy is a prerequisite for the reliability of the final results obtained by the combined procedure proposed in the present work. In the present study a new measure has been developed to utilize the metric distance obtained from cumulative distribution functions (cdfs). The measure has been evaluated for three different cases of distributions in order to assess the characteristics of the measure: The first case and the second are when the distribution is known as analytical distributions and the other case is when the distribution is unknown. The first case is given by symmetry analytical distributions. The second case consists of two asymmetry distributions of which the skewness is non zero

  1. Identifying complications of interventional procedures from UK routine healthcare databases: a systematic search for methods using clinical codes.

    Science.gov (United States)

    Keltie, Kim; Cole, Helen; Arber, Mick; Patrick, Hannah; Powell, John; Campbell, Bruce; Sims, Andrew

    2014-11-28

    Several authors have developed and applied methods to routine data sets to identify the nature and rate of complications following interventional procedures. But, to date, there has been no systematic search for such methods. The objective of this article was to find, classify and appraise published methods, based on analysis of clinical codes, which used routine healthcare databases in a United Kingdom setting to identify complications resulting from interventional procedures. A literature search strategy was developed to identify published studies that referred, in the title or abstract, to the name or acronym of a known routine healthcare database and to complications from procedures or devices. The following data sources were searched in February and March 2013: Cochrane Methods Register, Conference Proceedings Citation Index - Science, Econlit, EMBASE, Health Management Information Consortium, Health Technology Assessment database, MathSciNet, MEDLINE, MEDLINE in-process, OAIster, OpenGrey, Science Citation Index Expanded and ScienceDirect. Of the eligible papers, those which reported methods using clinical coding were classified and summarised in tabular form using the following headings: routine healthcare database; medical speciality; method for identifying complications; length of follow-up; method of recording comorbidity. The benefits and limitations of each approach were assessed. From 3688 papers identified from the literature search, 44 reported the use of clinical codes to identify complications, from which four distinct methods were identified: 1) searching the index admission for specified clinical codes, 2) searching a sequence of admissions for specified clinical codes, 3) searching for specified clinical codes for complications from procedures and devices within the International Classification of Diseases 10th revision (ICD-10) coding scheme which is the methodology recommended by NHS Classification Service, and 4) conducting manual clinical

  2. Experience with the Incomplete Cholesky Conjugate Gradient method in a diffusion code

    International Nuclear Information System (INIS)

    Hoebel, W.

    1985-01-01

    For the numerical solution of sparse systems of linear equations arising from finite difference approximation of the multidimensional neutron diffusion equation fast methods are needed. Effective algorithms for scalar computers may not be likewise suitable on vector computers. In the improved version DIXY2 of the Karlsruhe two-dimensional neutron diffusion code for rectangular geometries an Incomplete Cholesky Conjugate Gradient (ICCG) algorithm has been combined with the originally implemented Cyclically Reduced 4-Lines SOR (CR4LSOR) inner iteration method. The combined procedure is automatically activated for slowly converging applications, thus leading to a drastic reduction of iterations as well as CPU-times on a scalar computer. In a follow-up benchmark study necessary modifications to ICCG and CR4LSOR for their use on a vector computer were investigated. It was found that a modified preconditioning for the ICCG algorithm restricted to the block diagonal matrix is an effective method both on scalar and vector computers. With a splitting of the 9-band-matrix in two triangular Cholesky matrices necessary inversions are performed on a scalar machine by recursive forward and backward substitutions. On vector computers an additional factorization of the triangular matrices into four bidiagonal matrices enables Buneman reduction and the recursive inversion is restricted to a small system. A similar strategy can be realized with CR4LSOR if the unvectorizable Gauss-Seidel iteration is replaced by Double Jacobi and Buneman technique for a vector computer. Compared to single line blocking over the original mesh the cyclical 4-lines reduction of the DIXY inner iteration scheme reduces numbers of iterations and CPU-times considerably

  3. Experience with the incomplete Cholesky conjugate gradient method in a diffusion code

    International Nuclear Information System (INIS)

    Hoebel, W.

    1986-01-01

    For the numerical solution of sparse systems of linear equations arising from the finite difference approximation of the multidimensional neutron diffusion equation, fast methods are needed. Effective algorithms for scalar computers may not be likewise suitable on vector computers. In the improved version (DIXY2) of the Karlsruhe two-dimensional neutron diffusion code for rectangular geometries, an incomplete Cholesky conjugate gradient (ICCG) algorithm has been combined with the originally implemented cyclically reduced four-line successive overrelaxation (CR4LSOR) inner iteration method. The combined procedure is automatically activated for slowly converging applications, thus leading to a drastic reduction of iterations as well as CPU times on a scalar computer. In a follow-up benchmark study, necessary modifications to ICCG and CR4LSOR for use on a vector computer were investigated. It was found that a modified preconditioning for the ICCG algorithm restricted to the block diagonal matrix is an effective method both on scalar and vector computers. With a splitting of the nine-band matrix in two triangular Cholesky matrices, necessary inversions are performed on a scalar machine by recursive forward and backward substitutions. On vector computers an additional factorization of the triangular matrices into four bidiagonal matrices enables Buneman reduction, and the recursive inversion is restricted to a small system. A similar strategy can be realized with CR4LSOR if the unvectorizable Gauss-eidel iteration is replaced by Double Jacobi and Buneman techniques for a vector computer. Compared to single-line blocking over the original mesh, the cyclical four-line reduction of the DIXY inner iteration scheme reduces numbers of iterations and CPU times considerably

  4. The Use of Coding Methods to Estimate the Social Behavior Directed toward Peers and Adults of Preschoolers with ASD in TEACCH, LEAP, and Eclectic ''BAU'' Classrooms

    Science.gov (United States)

    Sam, Ann; Reszka, Stephanie; Odom, Samuel; Hume, Kara; Boyd, Brian

    2015-01-01

    Momentary time sampling, partial-interval recording, and event coding are observational coding methods commonly used to examine the social and challenging behaviors of children at risk for or with developmental delays or disabilities. Yet there is limited research comparing the accuracy of and relationship between these three coding methods. By…

  5. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  6. A massively parallel method of characteristic neutral particle transport code for GPUs

    International Nuclear Information System (INIS)

    Boyd, W. R.; Smith, K.; Forget, B.

    2013-01-01

    Over the past 20 years, parallel computing has enabled computers to grow ever larger and more powerful while scientific applications have advanced in sophistication and resolution. This trend is being challenged, however, as the power consumption for conventional parallel computing architectures has risen to unsustainable levels and memory limitations have come to dominate compute performance. Heterogeneous computing platforms, such as Graphics Processing Units (GPUs), are an increasingly popular paradigm for solving these issues. This paper explores the applicability of GPUs for deterministic neutron transport. A 2D method of characteristics (MOC) code - OpenMOC - has been developed with solvers for both shared memory multi-core platforms as well as GPUs. The multi-threading and memory locality methodologies for the GPU solver are presented. Performance results for the 2D C5G7 benchmark demonstrate 25-35 x speedup for MOC on the GPU. The lessons learned from this case study will provide the basis for further exploration of MOC on GPUs as well as design decisions for hardware vendors exploring technologies for the next generation of machines for scientific computing. (authors)

  7. A simple method for simulation of coherent synchrotron radiation in a tracking code

    International Nuclear Information System (INIS)

    Borland, M.

    2000-01-01

    Coherent synchrotron radiation (CSR) is of great interest to those designing accelerators as drivers for free-electron lasers (FELs). Although experimental evidence is incomplete, CSR is predicted to have potentially severe effects on the emittance of high-brightness electron beams. The performance of an FEL depends critically on the emittance, current, and energy spread of the beam. Attempts to increase the current through magnetic bunch compression can lead to increased emittance and energy spread due to CSR in the dipoles of such a compressor. The code elegant was used for design and simulation of the bunch compressor for the Low-Energy Undulator Test Line (LEUTL) FEL at the Advanced Photon Source (APS). In order to facilitate this design, a fast algorithm was developed based on the 1-D formalism of Saldin and coworkers. In addition, a plausible method of including CSR effects in drift spaces following the chicane magnets was developed and implemented. The algorithm is fast enough to permit running hundreds of tolerance simulations including CSR for 50 thousand particles. This article describes the details of the implementation and shows results for the APS bunch compressor

  8. Development of a code in three-dimensional cylindrical geometry based on analytic function expansion nodal (AFEN) method

    International Nuclear Information System (INIS)

    Lee, Joo Hee

    2006-02-01

    There is growing interest in developing pebble bed reactors (PBRs) as a candidate of very high temperature gas-cooled reactors (VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. But for realistic analysis of PBRs, there is strong desire of making available high fidelity nodal codes in three-dimensional (r,θ,z) cylindrical geometry. Recently, the Analytic Function Expansion Nodal (AFEN) method developed quite extensively in Cartesian (x,y,z) geometry and in hexagonal-z geometry was extended to two-group (r,z) cylindrical geometry, and gave very accurate results. In this thesis, we develop a method for the full three-dimensional cylindrical (r,θ,z) geometry and implement the method into a code named TOPS. The AFEN methodology in this geometry as in hexagonal geometry is 'robus' (e.g., no occurrence of singularity), due to the unique feature of the AFEN method that it does not use the transverse integration. The transverse integration in the usual nodal methods, however, leads to an impasse, that is, failure of the azimuthal term to be transverse-integrated over r-z surface. We use 13 nodal unknowns in an outer node and 7 nodal unknowns in an innermost node. The general solution of the node can be expressed in terms of that nodal unknowns, and can be updated using the nodal balance equation and the current continuity condition. For more realistic analysis of PBRs, we implemented em Marshak boundary condition to treat the incoming current zero boundary condition and the partial current translation (PCT) method to treat voids in the core. The TOPS code was verified in the various numerical tests derived from Dodds problem and PBMR-400 benchmark problem. The results of the TOPS code show high accuracy and fast computing time than the VENTURE code that is based on finite difference method (FDM)

  9. Computing travel time when the exact address is unknown: a comparison of point and polygon ZIP code approximation methods.

    Science.gov (United States)

    Berke, Ethan M; Shi, Xun

    2009-04-29

    Travel time is an important metric of geographic access to health care. We compared strategies of estimating travel times when only subject ZIP code data were available. Using simulated data from New Hampshire and Arizona, we estimated travel times to nearest cancer centers by using: 1) geometric centroid of ZIP code polygons as origins, 2) population centroids as origin, 3) service area rings around each cancer center, assigning subjects to rings by assuming they are evenly distributed within their ZIP code, 4) service area rings around each center, assuming the subjects follow the population distribution within the ZIP code. We used travel times based on street addresses as true values to validate estimates. Population-based methods have smaller errors than geometry-based methods. Within categories (geometry or population), centroid and service area methods have similar errors. Errors are smaller in urban areas than in rural areas. Population-based methods are superior to the geometry-based methods, with the population centroid method appearing to be the best choice for estimating travel time. Estimates in rural areas are less reliable.

  10. Pathway Detection from Protein Interaction Networks and Gene Expression Data Using Color-Coding Methods and A* Search Algorithms

    Directory of Open Access Journals (Sweden)

    Cheng-Yu Yeh

    2012-01-01

    Full Text Available With the large availability of protein interaction networks and microarray data supported, to identify the linear paths that have biological significance in search of a potential pathway is a challenge issue. We proposed a color-coding method based on the characteristics of biological network topology and applied heuristic search to speed up color-coding method. In the experiments, we tested our methods by applying to two datasets: yeast and human prostate cancer networks and gene expression data set. The comparisons of our method with other existing methods on known yeast MAPK pathways in terms of precision and recall show that we can find maximum number of the proteins and perform comparably well. On the other hand, our method is more efficient than previous ones and detects the paths of length 10 within 40 seconds using CPU Intel 1.73GHz and 1GB main memory running under windows operating system.

  11. FAFNER - a fully 3-D neutral beam injection code using Monte Carlo methods

    International Nuclear Information System (INIS)

    Lister, G.G.

    1985-01-01

    A computer code is described which models the injection of fast neutral particles into 3-dimensional toroidal plasmas and follows the paths of the resulting fast ions until they are either lost to the system or fully thermalised. A comprehensive model for the neutral beam injection system is included. The code is written especially for the use on the CRAY-1 computer: in particular, the modular nature of the program should enable the most time consuming sections of the program to be vectorised for each particular experiment to be modelled. The effects of plasma contamination by possible injection of impurities, such as oxygen, with the beams are also included. The code may also be readily adapted to plasmas for which a 1 or 2-dimensional description is adequate. It has also been constructed with a view to ready coupling with a transport or equilibrium code. (orig.)

  12. A Systematic Method for Verification and Validation of Gyrokinetic Microstability Codes

    Energy Technology Data Exchange (ETDEWEB)

    Bravenec, Ronald [Fourth State Research, Austin, TX (United States)

    2017-11-14

    My original proposal for the period Feb. 15, 2014 through Feb. 14, 2017 called for an integrated validation and verification effort carried out by myself with collaborators. The validation component would require experimental profile and power-balance analysis. In addition, it would require running the gyrokinetic codes varying the input profiles within experimental uncertainties to seek agreement with experiment before discounting a code as invalidated. Therefore, validation would require a major increase of effort over my previous grant periods which covered only code verification (code benchmarking). Consequently, I had requested full-time funding. Instead, I am being funded at somewhat less than half time (5 calendar months per year). As a consequence, I decided to forego the validation component and to only continue the verification efforts.

  13. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

  14. Coherent Synchrotron Radiation A Simulation Code Based on the Non-Linear Extension of the Operator Splitting Method

    CERN Document Server

    Dattoli, Giuseppe

    2005-01-01

    The coherent synchrotron radiation (CSR) is one of the main problems limiting the performance of high intensity electron accelerators. A code devoted to the analysis of this type of problems should be fast and reliable: conditions that are usually hardly achieved at the same time. In the past, codes based on Lie algebraic techniques have been very efficient to treat transport problem in accelerators. The extension of these method to the non-linear case is ideally suited to treat CSR instability problems. We report on the development of a numerical code, based on the solution of the Vlasov equation, with the inclusion of non-linear contribution due to wake field effects. The proposed solution method exploits an algebraic technique, using exponential operators implemented numerically in C++. We show that the integration procedure is capable of reproducing the onset of an instability and effects associated with bunching mechanisms leading to the growth of the instability itself. In addition, parametric studies a...

  15. Thermal hydraulic calculation of wire-wrapped bundles using a finite element method. Thesee code

    International Nuclear Information System (INIS)

    Rouzaud, P.; Gay, B.; Verviest, R.

    1981-07-01

    The physical and mathematical models used in the THESEE code now under development by the CEA/CEN Cadarache are presented. The objective of this code is to predict the fine three-dimensional temperature field in the sodium in a wire-wrapped rod bundle. Numerical results of THESEE are compared with measurements obtained by Belgonucleaire in 1976 in a sodium-cooled seven-rod bundle

  16. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  17. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  18. Review of hybrid (deterministic/Monte Carlo) radiation transport methods, codes, and applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, J.C.; Peplow, D.E.; Mosher, S.W.; Evans, T.M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications. (author)

  19. Development of an object oriented nodal code using the refined AFEN derived from the method of component decomposition

    International Nuclear Information System (INIS)

    Noh, J. M.; Yoo, J. W.; Joo, H. K.

    2004-01-01

    In this study, we invented a method of component decomposition to derive the systematic inter-nodal coupled equations of the refined AFEN method and developed an object oriented nodal code to solve the derived coupled equations. The method of component decomposition decomposes the intra-nodal flux expansion of a nodal method into even and odd components in three dimensions to reduce the large coupled linear system equation into several small single equations. This method requires no additional technique to accelerate the iteration process to solve the inter-nodal coupled equations, since the derived equations can automatically act as the coarse mesh re-balance equations. By utilizing the object oriented programming concepts such as abstraction, encapsulation, inheritance and polymorphism, dynamic memory allocation, and operator overloading, we developed an object oriented nodal code that can facilitate the input/output and the dynamic control of the memories, and can make the maintenance easy. (authors)

  20. PCR-free quantitative detection of genetically modified organism from raw materials. An electrochemiluminescence-based bio bar code method.

    Science.gov (United States)

    Zhu, Debin; Tang, Yabing; Xing, Da; Chen, Wei R

    2008-05-15

    A bio bar code assay based on oligonucleotide-modified gold nanoparticles (Au-NPs) provides a PCR-free method for quantitative detection of nucleic acid targets. However, the current bio bar code assay requires lengthy experimental procedures including the preparation and release of bar code DNA probes from the target-nanoparticle complex and immobilization and hybridization of the probes for quantification. Herein, we report a novel PCR-free electrochemiluminescence (ECL)-based bio bar code assay for the quantitative detection of genetically modified organism (GMO) from raw materials. It consists of tris-(2,2'-bipyridyl) ruthenium (TBR)-labeled bar code DNA, nucleic acid hybridization using Au-NPs and biotin-labeled probes, and selective capture of the hybridization complex by streptavidin-coated paramagnetic beads. The detection of target DNA is realized by direct measurement of ECL emission of TBR. It can quantitatively detect target nucleic acids with high speed and sensitivity. This method can be used to quantitatively detect GMO fragments from real GMO products.

  1. A modified carrier-to-code leveling method for retrieving ionospheric observables and detecting short-term temporal variability of receiver differential code biases

    Science.gov (United States)

    Zhang, Baocheng; Teunissen, Peter J. G.; Yuan, Yunbin; Zhang, Xiao; Li, Min

    2018-03-01

    Sensing the ionosphere with the global positioning system involves two sequential tasks, namely the ionospheric observable retrieval and the ionospheric parameter estimation. A prominent source of error has long been identified as short-term variability in receiver differential code bias (rDCB). We modify the carrier-to-code leveling (CCL), a method commonly used to accomplish the first task, through assuming rDCB to be unlinked in time. Aside from the ionospheric observables, which are affected by, among others, the rDCB at one reference epoch, the Modified CCL (MCCL) can also provide the rDCB offsets with respect to the reference epoch as by-products. Two consequences arise. First, MCCL is capable of excluding the effects of time-varying rDCB from the ionospheric observables, which, in turn, improves the quality of ionospheric parameters of interest. Second, MCCL has significant potential as a means to detect between-epoch fluctuations experienced by rDCB of a single receiver.

  2. Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4

    International Nuclear Information System (INIS)

    Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

    2004-01-01

    The expanding clinical use of low-energy photon emitting 125 I and 103 Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst ±5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately ±2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV

  3. Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4.

    Science.gov (United States)

    Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

    2004-02-07

    The expanding clinical use of low-energy photon emitting 125I and 103Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst +/- 5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately +/- 2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV.

  4. Comparison of the THYC and FLICA-3M codes by the pseudo-cubic thin-plate method; Comparaison par la methode des plaques de predicteurs de flux critique obtenus a l`aide des codes THYC et FLICA-3

    Energy Technology Data Exchange (ETDEWEB)

    Banner, D; Crecy, F de

    1993-06-01

    The pseudo cubic Spline method (PCSM) is a statistical tool developed by the CEA. It is designed to analyse experimental points and in particular thermalhydraulic data. Predictors of the occurrence of critical heat flux are obtained by using Spline functions. In this paper, predictors have been computed from the same CHF databases by using two different flow analyses to derive local thermal-hydraulic variables at the CHF location. In fact, CEA`s FLICA-3M represents rod bundles by interconnected subchannels whereas EDF`s THYC code uses a porous 3D approach. In a first step, the PCSM is briefly presented as well as the two codes studied here. Then, the comparison methodology is explained in order to prove that advanced analysis of thermalhydraulic codes can be achieved with the PCSM. (authors). 6 figs., 2 tabs., 5 refs.

  5. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  6. PLASTEF: a code for the numerical simulation of thermoelastoplastic behaviour of materials using the finite element method

    International Nuclear Information System (INIS)

    Basombrio, F.G.; Sanchez Sarmiento, G.

    1978-01-01

    A general code for solving two-dimensional thermo-elastoplastic problems in geometries of arbitrary shape using the finite element method, is presented. The initial stress incremental procedure was adopted, for given histories of load and temperature. Some classical applications are included. (Auth.)

  7. Acceleration of nodal diffusion code by Chebychev polynomial extrapolation method; Ubrzanje spoljasnjih iteracija difuzionog nodalnog proracuna Chebisevijevom ekstrapolacionom metodom

    Energy Technology Data Exchange (ETDEWEB)

    Zmijarevic, I; Tomashevic, Dj [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1988-07-01

    This paper presents Chebychev acceleration of outer iterations of a nodal diffusion code of high accuracy. Extrapolation parameters, unique for all moments are calculated using the node integrated distribution of fission source. Sample calculations are presented indicating the efficiency of method. (author)

  8. Calculation of extended shields in the Monte Carlo method using importance function (BRAND and DD code systems)

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Kagalenko, I.Eh.; Mironovich, Yu.N.

    1992-01-01

    Consideration is given of a technique and algorithms of constructing neutron trajectories in the Monte-Carlo method taking into account the data on adjoint transport equation solution. When simulating the transport part of transfer kernel the use is made of piecewise-linear approximation of free path length density along the particle motion direction. The approach has been implemented in programs within the framework of the BRAND code system. The importance is calculated in the multigroup P 1 -approximation within the framework of the DD-30 code system. The efficiency of the developed computation technique is demonstrated by means of solution of two model problems. 4 refs.; 2 tabs

  9. The response matrix method for the representation of the border conditions in the three-dimensional difussion codes

    International Nuclear Information System (INIS)

    Grant, C.R.

    1981-01-01

    It could take a considerable amount of memory and processing time to represent a reactor in its simulation by means of a diffusion code and considering areas in which nuclear and geometrical properties are invariant, such as reflector, water columns, etc. To avoid an explicit representation of these zones, a method employing a matrix was developed consisting in expressing the net currents of each group as a function of the total flux. Estimates are made for different geometries, introducing the PUMA difussion code of materials. Several tests made proved a very sound reliability of the results obtained in 2 and 5 groups. (author) [es

  10. Case file coding of child maltreatment: Methods, challenges, and innovations in a longitudinal project of youth in foster care☆

    Science.gov (United States)

    Huffhines, Lindsay; Tunno, Angela M.; Cho, Bridget; Hambrick, Erin P.; Campos, Ilse; Lichty, Brittany; Jackson, Yo

    2016-01-01

    State social service agency case files are a common mechanism for obtaining information about a child’s maltreatment history, yet these documents are often challenging for researchers to access, and then to process in a manner consistent with the requirements of social science research designs. Specifically, accessing and navigating case files is an extensive undertaking, and a task that many researchers have had to maneuver with little guidance. Even after the files are in hand and the research questions and relevant variables have been clarified, case file information about a child’s maltreatment exposure can be idiosyncratic, vague, inconsistent, and incomplete, making coding such information into useful variables for statistical analyses difficult. The Modified Maltreatment Classification System (MMCS) is a popular tool used to guide the process, and though comprehensive, this coding system cannot cover all idiosyncrasies found in case files. It is not clear from the literature how researchers implement this system while accounting for issues outside of the purview of the MMCS or that arise during MMCS use. Finally, a large yet reliable file coding team is essential to the process, however, the literature lacks training guidelines and methods for establishing reliability between coders. In an effort to move the field toward a common approach, the purpose of the present discussion is to detail the process used by one large-scale study of child maltreatment, the Studying Pathways to Adjustment and Resilience in Kids (SPARK) project, a longitudinal study of resilience in youth in foster care. The article addresses each phase of case file coding, from accessing case files, to identifying how to measure constructs of interest, to dealing with exceptions to the coding system, to coding variables reliably, to training large teams of coders and monitoring for fidelity. Implications for a comprehensive and efficient approach to case file coding are discussed. PMID

  11. Development of the system based code. v. 5. Method of margin exchange. pt. 2. Determination of quality assurance index based on a 'Vector Method'

    International Nuclear Information System (INIS)

    Asayama, Tai

    2003-03-01

    For the commercialization of fast breeder reactors, 'System Based Code', a completely new scheme of a code on structural integrity, is being developed. One of the distinguished features of the System Based Code is that it is able to determine a reasonable total margin on a structural of system, by allowing the exchanges of margins between various technical items. Detailed estimation of failure probability of a given combination of technical items and its comparison with a target value is one way to achieve this. However, simpler and easier methods that allow margin exchange without detailed calculation of failure probability are desirable in design. The authors have developed a simplified method such as a 'design factor method' from this viewpoint. This report describes a 'Vector Method', which was been newly developed. Following points are reported: 1) The Vector Method allows margin exchange evaluation on an 'equi-quality assurance plane' using vector calculation. Evaluation is easy and sufficient accuracy is achieved. The equi-quality assurance plane is obtained by a projection of an 'equi-failure probability surface in a n-dimensional space, which is calculated beforehand for typical combinations of design variables. 2) The Vector Method is considered to give the 'Quality Assurance Index Method' a probabilistic interpretation. 3) An algebraic method was proposed for the calculation of failure probabilities, which is necessary to obtain a equi-failure probability surface. This method calculates failure probabilities without using numerical methods such as Monte Carlo simulation or numerical integration. Under limited conditions, this method is quite effective compared to numerical methods. 4) An illustration of the procedure of margin exchange evaluation is given. It may be possible to use this method to optimize ISI plans; even it is not fully implemented in the System Based Code. (author)

  12. U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J.; Fanning, T. H.

    2017-06-26

    The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.

  13. Comparison of the THYC and FLICA-3M codes by the pseudo-cubic thin-plate method

    International Nuclear Information System (INIS)

    Banner, D.; Crecy, F. de.

    1993-06-01

    The pseudo cubic Spline method (PCSM) is a statistical tool developed by the CEA. It is designed to analyse experimental points and in particular thermalhydraulic data. Predictors of the occurrence of critical heat flux are obtained by using Spline functions. In this paper, predictors have been computed from the same CHF databases by using two different flow analyses to derive local thermal-hydraulic variables at the CHF location. In fact, CEA's FLICA-3M represents rod bundles by interconnected subchannels whereas EDF's THYC code uses a porous 3D approach. In a first step, the PCSM is briefly presented as well as the two codes studied here. Then, the comparison methodology is explained in order to prove that advanced analysis of thermalhydraulic codes can be achieved with the PCSM. (authors). 6 figs., 2 tabs., 5 refs

  14. Implementation of an implicit method into heat conduction calculation of TRAC-PF1/MOD2 code

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio

    1990-08-01

    A two-dimensional unsteady heat conduction equation is solved in the TRAC-PF/MOD2 code to calculate temperature transients in fuel rod. A large CPU time is often required to get stable solution of temperature transients in the TRAC calculation with a small axial node size (less than 1.0 mm), because the heat conduction equation is discretized explicitly. To eliminate the restriction of the maximum time step size by the heat conduction calculation, an implicit method for solving the heat condition equation was developed and implemented into the TRAC code. Several assessment calculations were performed with the original and modified TRAC codes. It is confirmed that the implicit method is reliable and is successfully implemented into the TRAC code through comparison with theoretical solutions and assessment calculation results. It is demonstrated that the implicit method makes the heat conduction calculation practical even for the analyses of temperature transients with the axial node size less than 0.1 mm. (author)

  15. Solution of charged particle transport equation by Monte-Carlo method in the BRANDZ code system

    International Nuclear Information System (INIS)

    Artamonov, S.N.; Androsenko, P.A.; Androsenko, A.A.

    1992-01-01

    Consideration is given to the issues of Monte-Carlo employment for the solution of charged particle transport equation and its implementation in the BRANDZ code system under the conditions of real 3D geometry and all the data available on radiation-to-matter interaction in multicomponent and multilayer targets. For the solution of implantation problem the results of BRANDZ data comparison with the experiments and calculations by other codes in complexes systems are presented. The results of direct nuclear pumping process simulation for laser-active media by a proton beam are also included. 4 refs.; 7 figs

  16. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

    International Nuclear Information System (INIS)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

    2003-01-01

    This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

  17. Relating system-to-CFD coupled code analyses to theoretical framework of a multi-scale method

    International Nuclear Information System (INIS)

    Cadinu, F.; Kozlowski, T.; Dinh, T.N.

    2007-01-01

    Over past decades, analyses of transient processes and accidents in a nuclear power plant have been performed, to a significant extent and with a great success, by means of so called system codes, e.g. RELAP5, CATHARE, ATHLET codes. These computer codes, based on a multi-fluid model of two-phase flow, provide an effective, one-dimensional description of the coolant thermal-hydraulics in the reactor system. For some components in the system, wherever needed, the effect of multi-dimensional flow is accounted for through approximate models. The later are derived from scaled experiments conducted for selected accident scenarios. Increasingly, however, we have to deal with newer and ever more complex accident scenarios. In some such cases the system codes fail to serve as simulation vehicle, largely due to its deficient treatment of multi-dimensional flow (in e.g. downcomer, lower plenum). A possible way of improvement is to use the techniques of Computational Fluid Dynamics (CFD). Based on solving Navier-Stokes equations, CFD codes have been developed and used, broadly, to perform analysis of multi-dimensional flow, dominantly in non-nuclear industry and for single-phase flow applications. It is clear that CFD simulations can not substitute system codes but just complement them. Given the intrinsic multi-scale nature of this problem, we propose to relate it to the more general field of research on multi-scale simulations. Even though multi-scale methods are developed on case-by-case basis, the need for a unified framework brought to the development of the heterogeneous multi-scale method (HMM)

  18. Video coding and decoding devices and methods preserving ppg relevant information

    NARCIS (Netherlands)

    2013-01-01

    The present invention relates to a video encoding device (10) for encoding video data and a corresponding video decoding device, wherein during decoding PPG relevant information shall be preserved. For this purpose the video coding device (10) comprises a first encoder (20) for encoding input video

  19. Development of flow network analysis code for block type VHTR core by linear theory method

    International Nuclear Information System (INIS)

    Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.

    2012-01-01

    VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

  20. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  1. Systematic analysis of coding and noncoding DNA sequences using methods of statistical linguistics

    Science.gov (United States)

    Mantegna, R. N.; Buldyrev, S. V.; Goldberger, A. L.; Havlin, S.; Peng, C. K.; Simons, M.; Stanley, H. E.

    1995-01-01

    We compare the statistical properties of coding and noncoding regions in eukaryotic and viral DNA sequences by adapting two tests developed for the analysis of natural languages and symbolic sequences. The data set comprises all 30 sequences of length above 50 000 base pairs in GenBank Release No. 81.0, as well as the recently published sequences of C. elegans chromosome III (2.2 Mbp) and yeast chromosome XI (661 Kbp). We find that for the three chromosomes we studied the statistical properties of noncoding regions appear to be closer to those observed in natural languages than those of coding regions. In particular, (i) a n-tuple Zipf analysis of noncoding regions reveals a regime close to power-law behavior while the coding regions show logarithmic behavior over a wide interval, while (ii) an n-gram entropy measurement shows that the noncoding regions have a lower n-gram entropy (and hence a larger "n-gram redundancy") than the coding regions. In contrast to the three chromosomes, we find that for vertebrates such as primates and rodents and for viral DNA, the difference between the statistical properties of coding and noncoding regions is not pronounced and therefore the results of the analyses of the investigated sequences are less conclusive. After noting the intrinsic limitations of the n-gram redundancy analysis, we also briefly discuss the failure of the zeroth- and first-order Markovian models or simple nucleotide repeats to account fully for these "linguistic" features of DNA. Finally, we emphasize that our results by no means prove the existence of a "language" in noncoding DNA.

  2. Development of a three-dimensional neutron transport code DFEM based on the double finite element method

    International Nuclear Information System (INIS)

    Fujimura, Toichiro

    1996-01-01

    A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author)

  3. Generalized concatenated quantum codes

    International Nuclear Information System (INIS)

    Grassl, Markus; Shor, Peter; Smith, Graeme; Smolin, John; Zeng Bei

    2009-01-01

    We discuss the concept of generalized concatenated quantum codes. This generalized concatenation method provides a systematical way for constructing good quantum codes, both stabilizer codes and nonadditive codes. Using this method, we construct families of single-error-correcting nonadditive quantum codes, in both binary and nonbinary cases, which not only outperform any stabilizer codes for finite block length but also asymptotically meet the quantum Hamming bound for large block length.

  4. Analysis of piping systems by finite element method using code SAP-IV

    International Nuclear Information System (INIS)

    Cizelj, L.; Ogrizek, D.

    1987-01-01

    Due to extensive and multiple use of the computer code SAP-IV we have decided to install it on VAX 11/750 machine. Installation required a large quantity of programming due to great discrepancies between the CDC (the original program version) and the VAX. Testing was performed basically in the field of pipe elements, based on a comparison between results obtained with the codes PSAFE2, DOCIJEV, PIPESD and SAP -V. Besides, the model of reactor pressure vessel with 3-D thick shell elements was done. The capabilities show good agreement with the results of other programs mentioned above. Along with the package installation, the graphical postprocessors being developed for mesh plotting. (author)

  5. Review of solution approach, methods, and recent results of the RELAP5 system code

    International Nuclear Information System (INIS)

    Trapp, J.A.; Ransom, V.H.

    1983-01-01

    The present RELAP5 code is based on a semi-implicit numerical scheme for the hydrodynamic model. The basic guidelines employed in the development of the semi-implicit numerical scheme are discussed and the numerical features of the scheme are illustrated by analysis for a simple, but analogous, single-equation model. The basic numerical scheme is recorded and results from several simulations are presented. The experimental results and code simulations are used in a complementary fashion to develop insights into nuclear-plant response that would not be obtained if either tool were used alone. Further analysis using the simple single-equation model is carried out to yield insights that are presently being used to implement a more-implicit multi-step scheme in the experimental version of RELAP5. The multi-step implicit scheme is also described

  6. Multidimensional method of spatially coupled approximation to the transverse escape in nodal codes

    International Nuclear Information System (INIS)

    Jatuff, F.E.

    1990-01-01

    A natural extension of the polynomic development programmed in RHENO code is presented, which adds to the variable order one-dimensional functions sum, a number of terms that represent functions of production. These new terms, which provide a direct determination of transverse escapes, are calculated from the new variables coupling among nodes: the 4 fluxes in rectangle vortices (bidimensional Cartesian geometry) or the 12 fluxes half-way through the parallelepiped edges (tridimensional Cartesian geometry). (Author) [es

  7. Method for computing self-consistent solution in a gun code

    Science.gov (United States)

    Nelson, Eric M

    2014-09-23

    Complex gun code computations can be made to converge more quickly based on a selection of one or more relaxation parameters. An eigenvalue analysis is applied to error residuals to identify two error eigenvalues that are associated with respective error residuals. Relaxation values can be selected based on these eigenvalues so that error residuals associated with each can be alternately reduced in successive iterations. In some examples, relaxation values that would be unstable if used alone can be used.

  8. Some questions of using coding theory and analytical calculation methods on computers

    International Nuclear Information System (INIS)

    Nikityuk, N.M.

    1987-01-01

    Main results of investigations devoted to the application of theory and practice of correcting codes are presented. These results are used to create very fast units for the selection of events registered in multichannel detectors of nuclear particles. Using this theory and analytical computing calculations, practically new combination devices, for example, parallel encoders, have been developed. Questions concerning the creation of a new algorithm for the calculation of digital functions by computers and problems of devising universal, dynamically reprogrammable logic modules are discussed

  9. The codes WAV3BDY and WAV4BDY and the variational Monte Carlo method

    International Nuclear Information System (INIS)

    Schiavilla, R.

    1987-01-01

    A description of the codes WAV3BDY and WAV4BDY, which generate the variational ground state wave functions of the A=3 and 4 nuclei, is given, followed by a discussion of the Monte Carlo integration technique, which is used to calculate expectation values and transition amplitudes of operators, and for whose implementation WAV3BDY and WAV4BDY are well suited

  10. Finite element methods in a simulation code for offshore wind turbines

    Science.gov (United States)

    Kurz, Wolfgang

    1994-06-01

    Offshore installation of wind turbines will become important for electricity supply in future. Wind conditions above sea are more favorable than on land and appropriate locations on land are limited and restricted. The dynamic behavior of advanced wind turbines is investigated with digital simulations to reduce time and cost in development and design phase. A wind turbine can be described and simulated as a multi-body system containing rigid and flexible bodies. Simulation of the non-linear motion of such a mechanical system using a multi-body system code is much faster than using a finite element code. However, a modal representation of the deformation field has to be incorporated in the multi-body system approach. The equations of motion of flexible bodies due to deformation are generated by finite element calculations. At Delft University of Technology the simulation code DUWECS has been developed which simulates the non-linear behavior of wind turbines in time domain. The wind turbine is divided in subcomponents which are represented by modules (e.g. rotor, tower etc.).

  11. MOCUM: A two-dimensional method of characteristics code based on constructive solid geometry and unstructured meshing for general geometries

    International Nuclear Information System (INIS)

    Yang Xue; Satvat, Nader

    2012-01-01

    Highlight: ► A two-dimensional numerical code based on the method of characteristics is developed. ► The complex arbitrary geometries are represented by constructive solid geometry and decomposed by unstructured meshing. ► Excellent agreement between Monte Carlo and the developed code is observed. ► High efficiency is achieved by parallel computing. - Abstract: A transport theory code MOCUM based on the method of characteristics as the flux solver with an advanced general geometry processor has been developed for two-dimensional rectangular and hexagonal lattice and full core neutronics modeling. In the code, the core structure is represented by the constructive solid geometry that uses regularized Boolean operations to build complex geometries from simple polygons. Arbitrary-precision arithmetic is also used in the process of building geometry objects to eliminate the round-off error from the commonly used double precision numbers. Then, the constructed core frame will be decomposed and refined into a Conforming Delaunay Triangulation to ensure the quality of the meshes. The code is fully parallelized using OpenMP and is verified and validated by various benchmarks representing rectangular, hexagonal, plate type and CANDU reactor geometries. Compared with Monte Carlo and deterministic reference solution, MOCUM results are highly accurate. The mentioned characteristics of the MOCUM make it a perfect tool for high fidelity full core calculation for current and GenIV reactor core designs. The detailed representation of reactor physics parameters can enhance the safety margins with acceptable confidence levels, which lead to more economically optimized designs.

  12. Pembuatan Kakas Pendeteksi Unused Method pada Kode Program PHP dengan Framework CodeIgniter Menggunakan Call Graph

    Directory of Open Access Journals (Sweden)

    Divi Galih Prasetyo Putri

    2014-03-01

    Full Text Available Proses evolusi dan perawatan dari sebuah sistem merupakan proses yang sangat penting dalam rekayasa perangkat lunak tidak terkecuali pada aplikasi web. Pada proses ini kebanyakan pengembang tidak lagi berpatokan pada rancangan sistem. Hal ini menyebabkan munculnya unused method. Bagian-bagian program ini tidak lagi terpakai namun masih berada dalam sistem. Keadaan ini meningkatkan kompleksitas dan mengurangi tingkat understandability sistem. Guna mendeteksi adanya unused method pada progam diperlukan teknik untuk melakukan code analysis. Teknik static analysis yang digunakan memanfaatkan call graph yang dibangun dari kode program untuk mengetahui adanya unused method. Call graph dibangun berdasarkan pemanggilan antar method. Aplikasi ini mendeteksi unused method pada kode program PHP yang dibangun menggunakan framework CodeIgniter. Kode program sebagai inputan diurai kedalam bentuk Abstract Syntax Tree (AST yang kemudian dimanfaatkan untuk melakukan analisis terhadap kode program. Proses analisis tersebut kemudian menghasilkan sebuah call graph. Dari call graph yang dihasilkan dapat dideteksi method-method mana saja yang tidak berhasil ditelusuri dan tergolong kedalam unused method. Kakas telah diuji coba pada 5 aplikasi PHP dengan hasil  rata-rata nilai presisi sistem sebesar 0.749 dan recall sebesar 1.

  13. Benchmark testing calculations for 232Th

    International Nuclear Information System (INIS)

    Liu Ping

    2003-01-01

    The cross sections of 232 Th from CNDC and JENDL-3.3 were processed with NJOY97.45 code in the ACE format for the continuous-energy Monte Carlo Code MCNP4C. The K eff values and central reaction rates based on CENDL-3.0, JENDL-3.3 and ENDF/B-6.2 were calculated using MCNP4C code for benchmark assembly, and the comparisons with experimental results are given. (author)

  14. Method for improvement of gamma-transition cascade spectra amplitude resolution by computer processing of coincidence codes

    International Nuclear Information System (INIS)

    Sukhovoj, A.M.; Khitrov, V.A.

    1982-01-01

    A method of improvement of amplitude resolution in the case of record of coinciding codes on the magnetic tape is suggested. It is shown on the record with Ge(Li) detectors of cascades of gamma-transitions from the 35 Cl(n, #betta#) reaction that total width at a half maximum of the peak may decrease by a factor of 2.6 for quanta with the energy similar to the neutron binding energy. Efficiency loss is absent

  15. Proceedings of the Seminar on Methods and Codes for Assessing the off-site consequences of nuclear accidents. Volume 1

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled 'methods for assessing the radiological impact of accidents' (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  16. Container for waste, identification code reading device thereof, method and system for controlling waste by using them

    International Nuclear Information System (INIS)

    Kikuchi, Takashi; Yoshida, Tomiji; Omote, Tatsuyuki.

    1991-01-01

    In the conventional method of controlling waste containers by labels attached thereto, the data relevant to wastes contained in the waste containers are limited. Further, if the label should be peeled off, there is a possibility that the wastes therein can no more be identified. Then, in the present invention, an identification plate is previously attached, to which mechanically readable codes or visually readable letters or numerical figures are written. Then, the identification codes can be read in a remote control manner at high speed and high reliability and the waste containers can be managed only by the identification codes of the containers. Further, the identification codes on the container are made so as to be free from aging degradation, thereby enabling to manage waste containers for long time storage. With such a constitution, since data can be inputted from an input terminal and a great amount of data such as concerning the source of wastes can be managed collectively on a software, the data can be managed easily. (T.M.)

  17. An Investigation of the Methods of Logicalizing the Code-Checking System for Architectural Design Review in New Taipei City

    Directory of Open Access Journals (Sweden)

    Wei-I Lee

    2016-12-01

    Full Text Available The New Taipei City Government developed a Code-checking System (CCS using Building Information Modeling (BIM technology to facilitate an architectural design review in 2014. This system was intended to solve problems caused by cognitive gaps between designer and reviewer in the design review process. Along with considering information technology, the most important issue for the system’s development has been the logicalization of literal building codes. Therefore, to enhance the reliability and performance of the CCS, this study uses the Fuzzy Delphi Method (FDM on the basis of design thinking and communication theory to investigate the semantic difference and cognitive gaps among participants in the design review process and to propose the direction of system development. Our empirical results lead us to recommend grouping multi-stage screening and weighted assisted logicalization of non-quantitative building codes to improve the operability of CCS. Furthermore, CCS should integrate the Expert Evaluation System (EES to evaluate the design value under qualitative building codes.

  18. Does health promotion need a Code of Ethics? Results from an IUHPE mixed method survey.

    Science.gov (United States)

    Bull, Torill; Riggs, Elisha; Nchogu, Sussy N

    2012-09-01

    Health promotion is an ethically challenging field involving constant reflection of values across multiple cultures of what is regarded as good and bad health promotion practice. While many disciplines are guided by a Code of Ethics (CoE) no such guide is available to health promoters. The International Union for Health Promotion and Education (IUHPE) has been nominated as a suitable candidate for developing such a code. It is within this context that the IUHPE Student and Early Career Network (ISECN), through its Ethics Working Group, has taken up the challenge of preparing the foundations for a CoE for health promotion. An online survey comprising open and closed-answer questions was used to gather the opinions of IUHPE members regarding the need for a CoE for health promotion. The quantitative data were calculated with descriptive analyses. A thematic analysis approach was used to analyze and interpret the qualitative data. IUHPE members (n = 236) from all global regions responded to the survey. The majority (52%) of the respondents had 11 years' experience or more in the field of health promotion. Ethical dilemmas were commonly encountered. The need for a CoE for health promotion was expressed by 83% of respondents. Respondents also offered their views of possibilities, ideas and challenges regarding the development of a CoE for health promotion. Considering that health promoters encounter ethical dilemmas frequently in their practice, this study reinforces the need to develop a CoE for the field. The recommendations from the survey provide a good basis for future work to develop such a code.

  19. Parallelization of one image compression method. Wavelet, Transform, Vector Quantization and Huffman Coding

    International Nuclear Information System (INIS)

    Moravie, Philippe

    1997-01-01

    Today, in the digitized satellite image domain, the needs for high dimension increase considerably. To transmit or to stock such images (more than 6000 by 6000 pixels), we need to reduce their data volume and so we have to use real-time image compression techniques. The large amount of computations required by image compression algorithms prohibits the use of common sequential processors, for the benefits of parallel computers. The study presented here deals with parallelization of a very efficient image compression scheme, based on three techniques: Wavelets Transform (WT), Vector Quantization (VQ) and Entropic Coding (EC). First, we studied and implemented the parallelism of each algorithm, in order to determine the architectural characteristics needed for real-time image compression. Then, we defined eight parallel architectures: 3 for Mallat algorithm (WT), 3 for Tree-Structured Vector Quantization (VQ) and 2 for Huffman Coding (EC). As our system has to be multi-purpose, we chose 3 global architectures between all of the 3x3x2 systems available. Because, for technological reasons, real-time is not reached at anytime (for all the compression parameter combinations), we also defined and evaluated two algorithmic optimizations: fix point precision and merging entropic coding in vector quantization. As a result, we defined a new multi-purpose multi-SMIMD parallel machine, able to compress digitized satellite image in real-time. The definition of the best suited architecture for real-time image compression was answered by presenting 3 parallel machines among which one multi-purpose, embedded and which might be used for other applications on board. (author) [fr

  20. New methods of analysis of materials strength data for the ASME Boiler and Pressure Vessel Code

    International Nuclear Information System (INIS)

    Booker, M.K.; Booker, B.L.P.

    1980-01-01

    Tensile and creep data of the type used to establish allowable stress levels for the ASME Boiler and Pressure Vessel Code have been examined for type 321H stainless steel. Both inhomogeneous, unbalanced data sets and well-planned homogeneous data sets have been examined. Data have been analyzed by implementing standard manual techniques on a modern digital computer. In addition, more sophisticated techniques, practical only through the use of the computer, have been applied. The result clearly demonstrates the efficacy of computerized techniques for these types of analyses

  1. Method to determine the strength of a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Chacon R, A.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Gallego, E.; Lorente, A. [Depto. Ingenieria Nuclear, Universidad Politecnica de Madrid, (Spain)

    2006-07-01

    The use of a gamma-ray spectrometer with a 3 {phi} x 3 NaI(Tl) detector, with a moderator sphere has been studied in the aim to measure the neutron fluence rate and to determine the source strength. Moderators with a large amount of hydrogen are able to slowdown and thermalize neutrons; once thermalized there is a probability that thermal neutron to be captured by hydrogen producing 2.22 MeV prompt gamma-ray. The pulse-height spectrum collected in a multicharmel analyzer shows a photopeak around 2.22 MeV whose net area is proportional to total neutron fluence rate and to the neutron source strength. The characteristics of this system were determined by a Monte Carlo study using the MCNP 4C code, where a detailed model of the Nal(Tl) was utilized. As moderators 3, 5, and 10 inches-diameter spheres where utilized and the response was calculated for monoenergetic and isotopic neutrons sources. (Author)

  2. A Simple Method for Static Load Balancing of Parallel FDTD Codes

    DEFF Research Database (Denmark)

    Franek, Ondrej

    2016-01-01

    A static method for balancing computational loads in parallel implementations of the finite-difference timedomain method is presented. The procedure is fairly straightforward and computationally inexpensive, thus providing an attractive alternative to optimization techniques. The method is descri...

  3. Prioritized Degree Distribution in Wireless Sensor Networks with a Network Coded Data Collection Method

    Science.gov (United States)

    Wan, Jan; Xiong, Naixue; Zhang, Wei; Zhang, Qinchao; Wan, Zheng

    2012-01-01

    The reliability of wireless sensor networks (WSNs) can be greatly affected by failures of sensor nodes due to energy exhaustion or the influence of brutal external environment conditions. Such failures seriously affect the data persistence and collection efficiency. Strategies based on network coding technology for WSNs such as LTCDS can improve the data persistence without mass redundancy. However, due to the bad intermediate performance of LTCDS, a serious ‘cliff effect’ may appear during the decoding period, and source data are hard to recover from sink nodes before sufficient encoded packets are collected. In this paper, the influence of coding degree distribution strategy on the ‘cliff effect’ is observed and the prioritized data storage and dissemination algorithm PLTD-ALPHA is presented to achieve better data persistence and recovering performance. With PLTD-ALPHA, the data in sensor network nodes present a trend that their degree distribution increases along with the degree level predefined, and the persistent data packets can be submitted to the sink node according to its degree in order. Finally, the performance of PLTD-ALPHA is evaluated and experiment results show that PLTD-ALPHA can greatly improve the data collection performance and decoding efficiency, while data persistence is not notably affected. PMID:23235451

  4. Prioritized degree distribution in wireless sensor networks with a network coded data collection method.

    Science.gov (United States)

    Wan, Jan; Xiong, Naixue; Zhang, Wei; Zhang, Qinchao; Wan, Zheng

    2012-12-12

    The reliability of wireless sensor networks (WSNs) can be greatly affected by failures of sensor nodes due to energy exhaustion or the influence of brutal external environment conditions. Such failures seriously affect the data persistence and collection efficiency. Strategies based on network coding technology for WSNs such as LTCDS can improve the data persistence without mass redundancy. However, due to the bad intermediate performance of LTCDS, a serious 'cliff effect' may appear during the decoding period, and source data are hard to recover from sink nodes before sufficient encoded packets are collected. In this paper, the influence of coding degree distribution strategy on the 'cliff effect' is observed and the prioritized data storage and dissemination algorithm PLTD-ALPHA is presented to achieve better data persistence and recovering performance. With PLTD-ALPHA, the data in sensor network nodes present a trend that their degree distribution increases along with the degree level predefined, and the persistent data packets can be submitted to the sink node according to its degree in order. Finally, the performance of PLTD-ALPHA is evaluated and experiment results show that PLTD-ALPHA can greatly improve the data collection performance and decoding efficiency, while data persistence is not notably affected.

  5. Computing observables in curved multifield models of inflation—A guide (with code) to the transport method

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Mafalda; Seery, David [Astronomy Centre, University of Sussex, Brighton BN1 9QH (United Kingdom); Frazer, Jonathan, E-mail: m.dias@sussex.ac.uk, E-mail: j.frazer@sussex.ac.uk, E-mail: a.liddle@sussex.ac.uk [Department of Theoretical Physics, University of the Basque Country, UPV/EHU, 48040 Bilbao (Spain)

    2015-12-01

    We describe how to apply the transport method to compute inflationary observables in a broad range of multiple-field models. The method is efficient and encompasses scenarios with curved field-space metrics, violations of slow-roll conditions and turns of the trajectory in field space. It can be used for an arbitrary mass spectrum, including massive modes and models with quasi-single-field dynamics. In this note we focus on practical issues. It is accompanied by a Mathematica code which can be used to explore suitable models, or as a basis for further development.

  6. Computing observables in curved multifield models of inflation—A guide (with code) to the transport method

    International Nuclear Information System (INIS)

    Dias, Mafalda; Seery, David; Frazer, Jonathan

    2015-01-01

    We describe how to apply the transport method to compute inflationary observables in a broad range of multiple-field models. The method is efficient and encompasses scenarios with curved field-space metrics, violations of slow-roll conditions and turns of the trajectory in field space. It can be used for an arbitrary mass spectrum, including massive modes and models with quasi-single-field dynamics. In this note we focus on practical issues. It is accompanied by a Mathematica code which can be used to explore suitable models, or as a basis for further development

  7. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  8. Pre-coding method and apparatus for multiple source or time-shifted single source data and corresponding inverse post-decoding method and apparatus

    Science.gov (United States)

    Yeh, Pen-Shu (Inventor)

    1998-01-01

    A pre-coding method and device for improving data compression performance by removing correlation between a first original data set and a second original data set, each having M members, respectively. The pre-coding method produces a compression-efficiency-enhancing double-difference data set. The method and device produce a double-difference data set, i.e., an adjacent-delta calculation performed on a cross-delta data set or a cross-delta calculation performed on two adjacent-delta data sets, from either one of (1) two adjacent spectral bands coming from two discrete sources, respectively, or (2) two time-shifted data sets coming from a single source. The resulting double-difference data set is then coded using either a distortionless data encoding scheme (entropy encoding) or a lossy data compression scheme. Also, a post-decoding method and device for recovering a second original data set having been represented by such a double-difference data set.

  9. HEFF---A user's manual and guide for the HEFF code for thermal-mechanical analysis using the boundary-element method

    International Nuclear Information System (INIS)

    St John, C.M.; Sanjeevan, K.

    1991-12-01

    The HEFF Code combines a simple boundary-element method of stress analysis with the closed form solutions for constant or exponentially decaying heat sources in an infinite elastic body to obtain an approximate method for analysis of underground excavations in a rock mass with heat generation. This manual describes the theoretical basis for the code, the code structure, model preparation, and step taken to assure that the code correctly performs its intended functions. The material contained within the report addresses the Software Quality Assurance Requirements for the Yucca Mountain Site Characterization Project. 13 refs., 26 figs., 14 tabs

  10. Entropy Coding in HEVC

    OpenAIRE

    Sze, Vivienne; Marpe, Detlev

    2014-01-01

    Context-Based Adaptive Binary Arithmetic Coding (CABAC) is a method of entropy coding first introduced in H.264/AVC and now used in the latest High Efficiency Video Coding (HEVC) standard. While it provides high coding efficiency, the data dependencies in H.264/AVC CABAC make it challenging to parallelize and thus limit its throughput. Accordingly, during the standardization of entropy coding for HEVC, both aspects of coding efficiency and throughput were considered. This chapter describes th...

  11. An Improved BeiDou-2 Satellite-Induced Code Bias Estimation Method

    Directory of Open Access Journals (Sweden)

    Jingyang Fu

    2018-04-01

    Full Text Available Different from GPS, GLONASS, GALILEO and BeiDou-3, it is confirmed that the code multipath bias (CMB, which originate from the satellite end and can be over 1 m, are commonly found in the code observations of BeiDou-2 (BDS IGSO and MEO satellites. In order to mitigate their adverse effects on absolute precise applications which use the code measurements, we propose in this paper an improved correction model to estimate the CMB. Different from the traditional model which considering the correction values are orbit-type dependent (estimating two sets of values for IGSO and MEO, respectively and modeling the CMB as a piecewise linear function with a elevation node separation of 10°, we estimate the corrections for each BDS IGSO + MEO satellite on one hand, and a denser elevation node separation of 5° is used to model the CMB variations on the other hand. Currently, the institutions such as IGS-MGEX operate over 120 stations which providing the daily BDS observations. These large amounts of data provide adequate support to refine the CMB estimation satellite by satellite in our improved model. One month BDS observations from MGEX are used for assessing the performance of the improved CMB model by means of precise point positioning (PPP. Experimental results show that for the satellites on the same orbit type, obvious differences can be found in the CMB at the same node and frequency. Results show that the new correction model can improve the wide-lane (WL ambiguity usage rate for WL fractional cycle bias estimation, shorten the WL and narrow-lane (NL time to first fix (TTFF in PPP ambiguity resolution (AR as well as improve the PPP positioning accuracy. With our improved correction model, the usage of WL ambiguity is increased from 94.1% to 96.0%, the WL and NL TTFF of PPP AR is shorten from 10.6 to 9.3 min, 67.9 to 63.3 min, respectively, compared with the traditional correction model. In addition, both the traditional and improved CMB model have

  12. Comparisons of ratchetting analysis methods using RCC-M, RCC-MR and ASME codes

    International Nuclear Information System (INIS)

    Yang Yu; Cabrillat, M.T.

    2005-01-01

    The present paper compares the simplified ratcheting analysis methods used in RCC-M, RCC-MR and ASME with some examples. Firstly, comparisons of the methods in RCC-M and efficiency diagram in RCC-MR are investigated. A special method is used to describe these two methods with curves in one coordinate, and the different conservation is demonstrated. RCC-M method is also be interpreted by SR (second ratio) and v (efficiency index) which is used in RCC-MR. Hence, we can easily compare the previous two methods by defining SR as abscissa and v as ordinate and plotting two curves of them. Secondly, comparisons of the efficiency curve in RCC-MR and methods in ASME-NH APPENDIX T are investigated, with significant creep. At last, two practical evaluations are performed to show the comparisons of aforementioned methods. (authors)

  13. Quasi-Newton methods for the acceleration of multi-physics codes

    CSIR Research Space (South Africa)

    Haelterman, R

    2017-08-01

    Full Text Available .E. Dennis, J.J. More´, Quasi-Newton methods: motivation and theory. SIAM Rev. 19, pp. 46–89 (1977) [11] J.E. Dennis, R.B. Schnabel, Least Change Secant Updates for quasi- Newton methods. SIAM Rev. 21, pp. 443–459 (1979) [12] G. Dhondt, CalculiX CrunchiX USER...) [25] J.M. Martinez, M.C. Zambaldi, An Inverse Column-Updating Method for solving large-scale nonlinear systems of equations. Optim. Methods Softw. 1, pp. 129–140 (1992) [26] J.M. Martinez, On the convergence of the column-updating method. Comp. Appl...

  14. Method of neutronic calculations for a spherical cell equivalent to cylindrical one for using computer codes in light water reactors in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Borges, V.; Sefidvash, F.; Rastogi, E.P.; Huria, H.C.; Krishnani, P.D.

    1989-01-01

    In order to use the existing light water reactor cell calculation codes for fluidized bed nuclear reactor having spherical fuel cells, an equivalence method has been developed. This method is shown to be adequate in calculation of the Dancoff factor. This method also was applicable in LEOPARD code and the results obtained in calculation of K ∞ was compared with the obtained using the DTF IV code, the results showed that the method is adequate for the calculations neutronics of the fluidized bed nuclear reactor. (author) [pt

  15. The development of high performance numerical simulation code for transient groundwater flow and reactive solute transport problems based on local discontinuous Galerkin method

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Motoshima, Takayuki; Naemura, Yumi; Kubo, Shin; Kanie, Shunji

    2009-01-01

    The authors develop a numerical code based on Local Discontinuous Galerkin Method for transient groundwater flow and reactive solute transport problems in order to make it possible to do three dimensional performance assessment on radioactive waste repositories at the earliest stage possible. Local discontinuous Galerkin Method is one of mixed finite element methods which are more accurate ones than standard finite element methods. In this paper, the developed numerical code is applied to several problems which are provided analytical solutions in order to examine its accuracy and flexibility. The results of the simulations show the new code gives highly accurate numeric solutions. (author)

  16. Monte Carlo method implemented in a finite element code with application to dynamic vacuum in particle accelerators

    CERN Document Server

    Garion, C

    2009-01-01

    Modern particle accelerators require UHV conditions during their operation. In the accelerating cavities, breakdowns can occur, releasing large amount of gas into the vacuum chamber. To determine the pressure profile along the cavity as a function of time, the time-dependent behaviour of the gas has to be simulated. To do that, it is useful to apply accurate three-dimensional method, such as Test Particles Monte Carlo. In this paper, a time-dependent Test Particles Monte Carlo is used. It has been implemented in a Finite Element code, CASTEM. The principle is to track a sample of molecules during time. The complex geometry of the cavities can be created either in the FE code or in a CAD software (CATIA in our case). The interface between the two softwares to export the geometry from CATIA to CASTEM is given. The algorithm of particle tracking for collisionless flow in the FE code is shown. Thermal outgassing, pumping surfaces and electron and/or ion stimulated desorption can all be generated as well as differ...

  17. Lifting scheme-based method for joint coding 3D stereo digital cinema with luminace correction and optimized prediction

    Science.gov (United States)

    Darazi, R.; Gouze, A.; Macq, B.

    2009-01-01

    Reproducing a natural and real scene as we see in the real world everyday is becoming more and more popular. Stereoscopic and multi-view techniques are used for this end. However due to the fact that more information are displayed requires supporting technologies such as digital compression to ensure the storage and transmission of the sequences. In this paper, a new scheme for stereo image coding is proposed. The original left and right images are jointly coded. The main idea is to optimally exploit the existing correlation between the two images. This is done by the design of an efficient transform that reduces the existing redundancy in the stereo image pair. This approach was inspired by Lifting Scheme (LS). The novelty in our work is that the prediction step is been replaced by an hybrid step that consists in disparity compensation followed by luminance correction and an optimized prediction step. The proposed scheme can be used for lossless and for lossy coding. Experimental results show improvement in terms of performance and complexity compared to recently proposed methods.

  18. A novel construction method of QC-LDPC codes based on the subgroup of the finite field multiplicative group for optical transmission systems

    Science.gov (United States)

    Yuan, Jian-guo; Zhou, Guang-xiang; Gao, Wen-chun; Wang, Yong; Lin, Jin-zhao; Pang, Yu

    2016-01-01

    According to the requirements of the increasing development for optical transmission systems, a novel construction method of quasi-cyclic low-density parity-check (QC-LDPC) codes based on the subgroup of the finite field multiplicative group is proposed. Furthermore, this construction method can effectively avoid the girth-4 phenomena and has the advantages such as simpler construction, easier implementation, lower encoding/decoding complexity, better girth properties and more flexible adjustment for the code length and code rate. The simulation results show that the error correction performance of the QC-LDPC(3 780,3 540) code with the code rate of 93.7% constructed by this proposed method is excellent, its net coding gain is respectively 0.3 dB, 0.55 dB, 1.4 dB and 1.98 dB higher than those of the QC-LDPC(5 334,4 962) code constructed by the method based on the inverse element characteristics in the finite field multiplicative group, the SCG-LDPC(3 969,3 720) code constructed by the systematically constructed Gallager (SCG) random construction method, the LDPC(32 640,30 592) code in ITU-T G.975.1 and the classic RS(255,239) code which is widely used in optical transmission systems in ITU-T G.975 at the bit error rate ( BER) of 10-7. Therefore, the constructed QC-LDPC(3 780,3 540) code is more suitable for optical transmission systems.

  19. Uncertainty Methods Framework Development for the TRACE Thermal-Hydraulics Code by the U.S.NRC

    International Nuclear Information System (INIS)

    Bajorek, Stephen M.; Gingrich, Chester

    2013-01-01

    The Code of Federal Regulations, Title 10, Part 50.46 requires that the Emergency Core Cooling System (ECCS) performance be evaluated for a number of postulated Loss-Of-Coolant-Accidents (LOCAs). The rule allows two methods for calculation of the acceptance criteria; using a realistic model in the so-called 'Best Estimate' approach, or the more prescriptive following Appendix K to Part 50. Because of the conservatism of Appendix K, recent Evaluation Model submittals to the NRC used the realistic approach. With this approach, the Evaluation Model must demonstrate that the Peak Cladding Temperature (PCT), the Maximum Local Oxidation (MLO) and Core-Wide Oxidation (CWO) remain below their regulatory limits with a 'high probability'. Guidance for Best Estimate calculations following 50.46(a)(1) was provided by Regulatory Guide 1.157. This Guide identified a 95% probability level as being acceptable for comparisons of best-estimate predictions to the applicable regulatory limits, but was vague with respect to acceptable methods in which to determine the code uncertainty. Nor, did it specify if a confidence level should be determined. As a result, vendors have developed Evaluation Models utilizing several different methods to combine uncertainty parameters and determine the PCT and other variables to a high probability. In order to quantify the accuracy of TRACE calculations for a wide variety of applications and to audit Best Estimate calculations made by industry, the NRC is developing its own independent methodology to determine the peak cladding temperature and other parameters of regulatory interest to a high probability. Because several methods are in use, and each vendor's methodology ranges different parameters, the NRC method must be flexible and sufficiently general. Not only must the method apply to LOCA analysis for conventional light-water reactors, it must also be extendable to new reactor designs and type of analyses where the acceptance criteria are less

  20. Advanced resonance self-shielding method for gray resonance treatment in lattice physics code GALAXY

    International Nuclear Information System (INIS)

    Koike, Hiroki; Yamaji, Kazuya; Kirimura, Kazuki; Sato, Daisuke; Matsumoto, Hideki; Yamamoto, Akio

    2012-01-01

    A new resonance self-shielding method based on the equivalence theory is developed for general application to the lattice physics calculations. The present scope includes commercial light water reactor (LWR) design applications which require both calculation accuracy and calculation speed. In order to develop the new method, all the calculation processes from cross-section library preparation to effective cross-section generation are reviewed and reframed by adopting the current enhanced methodologies for lattice calculations. The new method is composed of the following four key methods: (1) cross-section library generation method with a polynomial hyperbolic tangent formulation, (2) resonance self-shielding method based on the multi-term rational approximation for general lattice geometry and gray resonance absorbers, (3) spatially dependent gray resonance self-shielding method for generation of intra-pellet power profile and (4) integrated reaction rate preservation method between the multi-group and the ultra-fine-group calculations. From the various verifications and validations, applicability of the present resonance treatment is totally confirmed. As a result, the new resonance self-shielding method is established, not only by extension of a past concentrated effort in the reactor physics research field, but also by unification of newly developed unique and challenging techniques for practical application to the lattice physics calculations. (author)

  1. Parallel deposition, sorting, and reordering methods in the Hybrid Ordered Plasma Simulation (HOPS) code

    International Nuclear Information System (INIS)

    Anderson, D.V.; Shumaker, D.E.

    1993-01-01

    From a computational standpoint, particle simulation calculations for plasmas have not adapted well to the transitions from scalar to vector processing nor from serial to parallel environments. They have suffered from inordinate and excessive accessing of computer memory and have been hobbled by relatively inefficient gather-scatter constructs resulting from the use of indirect indexing. Lastly, the many-to-one mapping characteristic of the deposition phase has made it difficult to perform this in parallel. The authors' code sorts and reorders the particles in a spatial order. This allows them to greatly reduce the memory references, to run in directly indexed vector mode, and to employ domain decomposition to achieve parallelization. In this hybrid simulation the electrons are modeled as a fluid and the field equations solved are obtained from the electron momentum equation together with the pre-Maxwell equations (displacement current neglected). Either zero or finite electron mass can be used in the electron model. The resulting field equations are solved with an iteratively explicit procedure which is thus trivial to parallelize. Likewise, the field interpolations and the particle pushing is simple to parallelize. The deposition, sorting, and reordering phases are less simple and it is for these that the authors present detailed algorithms. They have now successfully tested the parallel version of HOPS in serial mode and it is now being readied for parallel execution on the Cray C-90. They will then port HOPS to a massively parallel computer, in the next year

  2. Method of accounting for code safety valve setpoint drift in safety analyses

    International Nuclear Information System (INIS)

    Rousseau, K.R.; Bergeron, P.A.

    1989-01-01

    In performing the safety analyses for transients that result in a challenge to the reactor coolant system (RCS) pressure boundary, the general acceptance criterion is that the peak RCS pressure not exceed the American Society of Mechanical Engineers limit of 110% of the design pressure. Without crediting non-safety-grade pressure mitigating systems, protection from this limit is mainly provided by the primary and secondary code safety valves. In theory, the combination of relief capacity and setpoints for these valves is designed to provide this protection. Generally, banks of valves are set at varying setpoints staggered by 15- to 20-psid increments to minimize the number of valves that would open by an overpressure challenge. In practice, however, when these valves are removed and tested (typically during a refueling outage), setpoints are sometimes found to have drifted by >50 psid. This drift should be accounted for during the performance of the safety analysis. This paper describes analyses performed by Yankee Atomic Electric Company (YAEC) to account for setpoint drift in safety valves from testing. The results of these analyses are used to define safety valve operability or acceptance criteria

  3. Radiation field characterization of a BNCT research facility using Monte Carlo method - code MCNP-4B

    International Nuclear Information System (INIS)

    Hernandez, Antonio Carlos

    2002-01-01

    Boron Neutron Capture Therapy - BNCT - is a selective cancer treatment and arises as an alternative therapy to treat cancer when usual techniques - surgery, chemotherapy or radiotherapy - show no satisfactory results. The main proposal of this work is to project a facility to BNCT studies. This facility relies on the use of an Am Be neutron source and on a set of moderators, filters and shielding which will provide the best neutron/gamma beam characteristic for these Becton studies, i.e., high intensity thermal and/or epithermal neutron fluxes and with the minimum feasible gamma rays and fast neutrons contaminants. A computational model of the experiment was used to obtain the radiation field in the sample irradiation position. The calculations have been performed with the MCNP 4B Monte Carlo Code and the results obtained can be regarded as satisfactory, i.e., a thermal neutron fluencyN T = 1,35x10 8 n/cm , a fast neutron dose of 5,86x10 -10 Gy/N T and a gamma ray dose of 8,30x10 -14 Gy/N T . (author)

  4. Radiation field characterization of a BNCT research facility using Monte Carlo Method - Code MCNP-4B

    International Nuclear Information System (INIS)

    Hernandes, Antonio Carlos

    2002-01-01

    Boron Neutron Capture Therapy - BNCT- is a selective cancer treatment and arises as an alternative therapy to treat cancer when usual techniques - surgery, chemotherapy or radiotherapy - show no satisfactory results. The main proposal of this work is to project a facility to BNCT studies. This facility relies on the use of an AmBe neutron source and on a set of moderators, filters and shielding which will provide the best neutron/gamma beam characteristic for these BNCT studies, i.e., high intensity thermal and/or epithermal neutron fluxes and with the minimum feasible gamma rays and fast neutrons contaminants. A computational model of the experiment was used to obtain the radiation field in the sample irradiation position. The calculations have been performed with the MCNP 4B Monte Carlo Code and the results obtained can be regarded as satisfactory, i.e., a thermal neutron fluency Ν Τ = 1,35x10 8 n/cm 2 , a fast neutron dose of 5,86x -1 0 Gy/Ν Τ and a gamma ray dose of 8,30x -14 Gy/Ν Τ . (author)

  5. A research on verification of the CONTAIN CODE model and the uncertainty reduction method for containment integrity

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae-Hong; Kim, Moo-Hwan; Bae, Seong-Won; Byun, Sang-Chul [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1998-03-15

    The final objectives of this study are to establish the way of measuring the integrity of containment building structures and safety analysis in the period of a postuIated severe accidents and to decrease the uncertainty of these methods. For that object, the CONTAIN 1.2 codes model for analyzing the severe accidents phenomena and the heat transfer between the air inside the containment buildings and inner walls have been reviewed and analyzed. For the double containment wall provided to the next generation nuclear reactor, which is different to the previous type of containment, the temperature and pressure rising history were calculated and compared to the results of previous ones.

  6. A new phase coding method using a slice selection gradient for high speed flow velocity meaurements in NMR tomography

    International Nuclear Information System (INIS)

    Oh, C.H.; Cho, Z.H.; California Univ., Irvine

    1986-01-01

    A new phase coding method using a selection gradient for high speed NMR flow velocity measurements is introduced and discussed. To establish a phase-velocity relationship of flow under the slice selection gradient and spin-echo RF pulse, the Bloch equation was numerically solved under the assumption that only one directional flow exists, i.e. in the direction of slice selection. Details of the numerical solution of the Bloch equation and techniques related to the numerical computations are also given. Finally, using the numerical calculation, high speed flow velocity measurement was attempted and found to be in good agreement with other complementary controlled measurements. (author)

  7. Video coding and decoding devices and methods preserving PPG relevant information

    NARCIS (Netherlands)

    2015-01-01

    The present invention relates to a video encoding device (10, 10', 10") and method for encoding video data and to a corresponding video decoding device (60, 60') and method. To preserve PPG relevant information after encoding without requiring a large amount of additional data for the video encoder

  8. Video coding and decoding devices and methods preserving ppg relevant information

    NARCIS (Netherlands)

    2013-01-01

    The present invention relates to a video encoding device (10, 10', 10'') and method for encoding video data and to a corresponding video decoding device (60, 60') and method. To preserve PPG relevant information after encoding without requiring a large amount of additional data for the video encoder

  9. The probabilistic method of WWER fuel rod strength estimation using the START-3 code

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Medvedev, A.V.; Bogatyr, S.M.; Sokolov, F.F.; Khramtsov, M.V.

    2001-01-01

    During the last years probability methods of studying were widely used to determine the influence exerted by the geometry, technology and performance parameters of a fuel rod on the characteristics of its condition. Despite the diversity of probability methods their basis is formed by the simplest schema of the Monte-Carlo method (MC). This schema assumes a great number of the realizations of a random value and the statistical assessment of its characteristics. To generate random values, use is usually made of a pseudo-random number generator. The application of the quasi-random sequence elements in place of the latter substantially reduces the machine time since it promotes a quicker convergence of the method. Probability methods used to study the characteristics of a fuel rod condition can be considered to be an auxiliary means of deterministic calculations that allows the assessment of the conservatism degree of design calculations. (author)

  10. Measuring the implementation of codes of conduct. An assessment method based on a process approach of the responsible organisation

    NARCIS (Netherlands)

    Nijhof, A.H.J.; Cludts, Stephan; Fisscher, O.A.M.; Laan, Albertus

    2003-01-01

    More and more organisations formulate a code of conduct in order to stimulate responsible behaviour among their members. Much time and energy is usually spent fixing the content of the code but many organisations get stuck in the challenge of implementing and maintaining the code. The code then

  11. Using of the Serpent code based on the Monte-Carlo method for calculation of the VVER-1000 fuel assembly characteristics

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2016-12-01

    Full Text Available The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc., which are used for neutron-physical analysis of VVER type fuel, are presented.

  12. Methods for Ensuring High Quality of Coding of Cause of Death. The Mortality Register to Follow Southern Urals Populations Exposed to Radiation.

    Science.gov (United States)

    Startsev, N; Dimov, P; Grosche, B; Tretyakov, F; Schüz, J; Akleyev, A

    2015-01-01

    To follow up populations exposed to several radiation accidents in the Southern Urals, a cause-of-death registry was established at the Urals Center capturing deaths in the Chelyabinsk, Kurgan and Sverdlovsk region since 1950. When registering deaths over such a long time period, quality measures need to be in place to maintain quality and reduce the impact of individual coders as well as quality changes in death certificates. To ensure the uniformity of coding, a method for semi-automatic coding was developed, which is described here. Briefly, the method is based on a dynamic thesaurus, database-supported coding and parallel coding by two different individuals. A comparison of the proposed method for organizing the coding process with the common procedure of coding showed good agreement, with, at the end of the coding process, 70  - 90% agreement for the three-digit ICD -9 rubrics. The semi-automatic method ensures a sufficiently high quality of coding by at the same time providing an opportunity to reduce the labor intensity inherent in the creation of large-volume cause-of-death registries.

  13. MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

    International Nuclear Information System (INIS)

    2005-01-01

    A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

  14. Estimation of POL-iteration methods in fast running DNBR code

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Kim, S. J.; Seo, K. W.; Hwang, D. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, various root finding methods are applied to the POL-iteration module in SCOMS and POLiteration efficiency is compared with reference method. On the base of these results, optimum algorithm of POL iteration is selected. The POL requires the iteration until present local power reach limit power. The process to search the limiting power is equivalent with a root finding of nonlinear equation. POL iteration process involved in online monitoring system used a variant bisection method that is the most robust algorithm to find the root of nonlinear equation. The method including the interval accelerating factor and escaping routine out of ill-posed condition assured the robustness of SCOMS system. POL iteration module in SCOMS shall satisfy the requirement which is a minimum calculation time. For this requirement of calculation time, non-iterative algorithm, few channel model, simple steam table are implemented into SCOMS to improve the calculation time. MDNBR evaluation at a given operating condition requires the DNBR calculation at all axial locations. An increasing of POL-iteration number increased a calculation load of SCOMS significantly. Therefore, calculation efficiency of SCOMS is strongly dependent on the POL iteration number. In case study, the iterations of the methods have a superlinear convergence for finding limiting power but Brent method shows a quardratic convergence speed. These methods are effective and better than the reference bisection algorithm.

  15. A comparison of different quasi-newton acceleration methods for partitioned multi-physics codes

    CSIR Research Space (South Africa)

    Haelterman, R

    2018-02-01

    Full Text Available & structures, 88/7, pp. 446–457 (2010) 8. J.E. Dennis, J.J. More´, Quasi-Newton methods: motivation and theory. SIAM Rev. 19, pp. 46–89 (1977) A Comparison of Quasi-Newton Acceleration Methods 15 9. J.E. Dennis, R.B. Schnabel, Least Change Secant Updates... Dois Metodos de Broyden. Mat. Apl. Comput. 1/2, pp. 135– 143 (1982) 25. J.M. Martinez, A quasi-Newton method with modification of one column per iteration. Com- puting 33, pp. 353–362 (1984) 26. J.M. Martinez, M.C. Zambaldi, An Inverse Column...

  16. Method for improving the gamma-transition cascade spectra amplitude resolution during coincidence code computerized processing

    International Nuclear Information System (INIS)

    Sukhovoj, A.M.; Khitrov, V.A.

    1984-01-01

    A method of unfolding the differential γ-cascade spectra during radiation capture of slow neutrons based on the computeri-- zed processing of the results of measurements performed, by means of a spectrometer with two Ge(Li) detectors is suggested. The efficiency of the method is illustrated using as an example the spectrum of 35 Cl(n, γ) reaction corresponding to the 8580 keV peak. It is shown that the above approach permits to improve the resolution by 1.2-2.6 times without decrease in registration efficiency within the framework of the method of coincidence pulse amplitude summation

  17. Parallel implementation of a dynamic unstructured chimera method in the DLR finite volume TAU-code

    International Nuclear Information System (INIS)

    Madrane, A.; Raichle, A.; Stuermer, A.

    2004-01-01

    Aerodynamic problems involving moving geometries have many applications, including store separation, high-speed train entering into a tunnel, simulation of full configurations of the helicopter and fast maneuverability. Overset grid method offers the option of calculating these procedures. The solution process uses a grid system that discretizes the problem domain by using separately generated but overlapping unstructured grids that update and exchange boundary information through interpolation. However, such computations are complicated and time consuming. Parallel computing offers a very effective way to improve the productivity in doing computational fluid dynamics (CFD). Therefore the purpose of this study is to develop an efficient parallel computation algorithm for analyzing the flowfield of complex geometries using overset grids method. The strategy adopted in the parallelization of the overset grids method including the use of data structures and communication, is described. Numerical results are presented to demonstrate the efficiency of the resulting parallel overset grids method. (author)

  18. Parallel implementation of a dynamic unstructured chimera method in the DLR finite volume TAU-code

    Energy Technology Data Exchange (ETDEWEB)

    Madrane, A.; Raichle, A.; Stuermer, A. [German Aerospace Center, DLR, Numerical Methods, Inst. of Aerodynamics and Flow Technology, Braunschweig (Germany)]. E-mail: aziz.madrane@dlr.de

    2004-07-01

    Aerodynamic problems involving moving geometries have many applications, including store separation, high-speed train entering into a tunnel, simulation of full configurations of the helicopter and fast maneuverability. Overset grid method offers the option of calculating these procedures. The solution process uses a grid system that discretizes the problem domain by using separately generated but overlapping unstructured grids that update and exchange boundary information through interpolation. However, such computations are complicated and time consuming. Parallel computing offers a very effective way to improve the productivity in doing computational fluid dynamics (CFD). Therefore the purpose of this study is to develop an efficient parallel computation algorithm for analyzing the flowfield of complex geometries using overset grids method. The strategy adopted in the parallelization of the overset grids method including the use of data structures and communication, is described. Numerical results are presented to demonstrate the efficiency of the resulting parallel overset grids method. (author)

  19. Summary of the models and methods for the FEHM application - a finite-element heat- and mass-transfer code

    International Nuclear Information System (INIS)

    Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.

    1997-07-01

    The mathematical models and numerical methods employed by the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multi-component flow in porous media, are described. The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The component models of FEHM are discussed. The first major component, Flow- and Energy-Transport Equations, deals with heat conduction; heat and mass transfer with pressure- and temperature-dependent properties, relative permeabilities and capillary pressures; isothermal air-water transport; and heat and mass transfer with noncondensible gas. The second component, Dual-Porosity and Double-Porosity/Double-Permeability Formulation, is designed for problems dominated by fracture flow. Another component, The Solute-Transport Models, includes both a reactive-transport model that simulates transport of multiple solutes with chemical reaction and a particle-tracking model. Finally, the component, Constitutive Relationships, deals with pressure- and temperature-dependent fluid/air/gas properties, relative permeabilities and capillary pressures, stress dependencies, and reactive and sorbing solutes. Each of these components is discussed in detail, including purpose, assumptions and limitations, derivation, applications, numerical method type, derivation of numerical model, location in the FEHM code flow, numerical stability and accuracy, and alternative approaches to modeling the component

  20. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  1. An Optimal Linear Coding for Index Coding Problem

    OpenAIRE

    Pezeshkpour, Pouya

    2015-01-01

    An optimal linear coding solution for index coding problem is established. Instead of network coding approach by focus on graph theoric and algebraic methods a linear coding program for solving both unicast and groupcast index coding problem is presented. The coding is proved to be the optimal solution from the linear perspective and can be easily utilize for any number of messages. The importance of this work is lying mostly on the usage of the presented coding in the groupcast index coding ...

  2. Proposal of flexible atomic and molecular process management for Monte Carlo impurity transport code based on object oriented method

    International Nuclear Information System (INIS)

    Asano, K.; Ohno, N.; Takamura, S.

    2001-01-01

    Monte Carlo simulation code on impurity transport has been developed by several groups to be utilized mainly for fusion related edge plasmas. State of impurity particle is determined by atomic and molecular processes such as ionization, charge exchange in plasma. A lot of atomic and molecular processes have been considered because the edge plasma has not only impurity atoms, but also impurity molecules mainly related to chemical erosion of carbon materials, and their cross sections have been given experimentally and theoretically. We need to reveal which process is essential in a given edge plasma condition. Monte Carlo simulation code, which takes such various atomic and molecular processes into account, is necessary to investigate the behavior of impurity particle in plasmas. Usually, the impurity transport simulation code has been intended for some specific atomic and molecular processes so that the introduction of a new process forces complicated programming work. In order to evaluate various proposed atomic and molecular processes, a flexible management of atomic and molecular reaction should be established. We have developed the impurity transport simulation code based on object-oriented method. By employing object-oriented programming, we can handle each particle as 'object', which enfolds data and procedure function itself. A user (notice, not programmer) can define property of each particle species and the related atomic and molecular processes and then each 'object' is defined by analyzing this information. According to the relation among plasma particle species, objects are connected with each other and change their state by themselves. Dynamic allocation of these objects to program memory is employed to adapt for arbitrary number of species and atomic/molecular reactions. Thus we can treat arbitrary species and process starting from, for instance, methane and acetylene. Such a software procedure would be useful also for industrial application plasmas

  3. Benchmarking of EPRI-cell epithermal methods with the point-energy discrete-ordinates code (OZMA)

    International Nuclear Information System (INIS)

    Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.

    1982-01-01

    The purpose of the present study is to benchmark E-C resonance-shielding and cell-averaging methods against a rigorous deterministic solution on a fine-group level (approx. 30 groups between 1 eV and 5.5 keV). The benchmark code used is OZMA, which solves the space-dependent slowing-down equations using continuous-energy discrete ordinates or integral transport theory to produce fine-group cross sections. Results are given for three water-moderated lattices - a mixed oxide, a uranium method, and a tight-pitch high-conversion uranium oxide configuration. The latter two lattices were chosen because of the strong self shielding of the 238 U resonances

  4. Using lattice tools and unfolding methods for hpge detector efficiency simulation with the Monte Carlo code MCNP5

    International Nuclear Information System (INIS)

    Querol, A.; Gallardo, S.; Ródenas, J.; Verdú, G.

    2015-01-01

    In environmental radioactivity measurements, High Purity Germanium (HPGe) detectors are commonly used due to their excellent resolution. Efficiency calibration of detectors is essential to determine activity of radionuclides. The Monte Carlo method has been proved to be a powerful tool to complement efficiency calculations. In aged detectors, efficiency is partially deteriorated due to the dead layer increasing and consequently, the active volume decreasing. The characterization of the radiation transport in the dead layer is essential for a realistic HPGe simulation. In this work, the MCNP5 code is used to calculate the detector efficiency. The F4MESH tally is used to determine the photon and electron fluence in the dead layer and the active volume. The energy deposited in the Ge has been analyzed using the ⁎F8 tally. The F8 tally is used to obtain spectra and to calculate the detector efficiency. When the photon fluence and the energy deposition in the crystal are known, some unfolding methods can be used to estimate the activity of a given source. In this way, the efficiency is obtained and serves to verify the value obtained by other methods. - Highlights: • The MCNP5 code is used to estimate the dead layer thickness of an HPGe detector. • The F4MESH tally is applied to verify where interactions occur into the Ge crystal. • PHD and the energy deposited are obtained with F8 and ⁎F8 tallies, respectively. • An average dead layer between 70 and 80 µm is obtained for the HPGe studied. • The efficiency is calculated applying the TSVD method to the response matrix.

  5. R and D on automatic modeling methods for Monte Carlo codes FLUKA

    International Nuclear Information System (INIS)

    Wang Dianxi; Hu Liqin; Wang Guozhong; Zhao Zijia; Nie Fanzhi; Wu Yican; Long Pengcheng

    2013-01-01

    FLUKA is a fully integrated particle physics Monte Carlo simulation package. It is necessary to create the geometry models before calculation. However, it is time- consuming and error-prone to describe the geometry models manually. This study developed an automatic modeling method which could automatically convert computer-aided design (CAD) geometry models into FLUKA models. The conversion program was integrated into CAD/image-based automatic modeling program for nuclear and radiation transport simulation (MCAM). Its correctness has been demonstrated. (authors)

  6. Computational methods and implementation of the 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction

    International Nuclear Information System (INIS)

    Aragones, J.M.; Ahnert, C.

    1995-01-01

    New computational methods have been developed in our 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction. They improve the accuracy and efficiency of the coupled neutronic-thermalhydraulic solution and extend its scope to provide, mainly, the calculation of: the fission reaction rates at the incore mini-detectors; the responses at the excore detectors (power range); the temperatures at the thermocouple locations; and the in-vessel distribution of the loop cold-leg inlet coolant conditions in the reflector and core channels, and to the hot-leg outlets per loop. The functional capabilities implemented in the extended SIMTRAN code for online utilization include: online surveillance, incore-excore calibration, evaluation of peak power factors and thermal margins, nominal update and cycle follow, prediction of maneuvers and diagnosis of fast transients and oscillations. The new code has been installed at the Vandellos-II PWR unit in Spain, since the startup of its cycle 7 in mid-June, 1994. The computational implementation has been performed on HP-700 workstations under the HP-UX Unix system, including the machine-man interfaces for online acquisition of measured data and interactive graphical utilization, in C and X11. The agreement of the simulated results with the measured data, during the startup tests and first months of actual operation, is well within the accuracy requirements. The performance and usefulness shown during the testing and demo phase, to be extended along this cycle, has proved that SIMTRAN and the man-machine graphic user interface have the qualities for a fast, accurate, user friendly, reliable, detailed and comprehensive online core surveillance and prediction

  7. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  8. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

    2003-07-01

    This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

  9. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  10. Assessment of gamma irradiation heating and damage in miniature neutron source reactor vessel using computational methods and SRIM - TRIM code

    International Nuclear Information System (INIS)

    Appiah-Ofori, F. F.

    2014-07-01

    The Effects of Gamma Radiation Heating and Irradiation Damage in the reactor vessel of Ghana Research Reactor 1, Miniature Neutron Source Reactor were assessed using Implicit Control Volume Finite Difference Numerical Computation and validated by SRIM - TRIM Code. It was assumed that 5.0 MeV of gamma rays from the reactor core generate heat which interact and absorbed completely by the interior surface of the MNSR vessel which affects it performance due to the induced displacement damage. This displacement damage is as result of lattice defects being created which impair the vessel through formation of point defect clusters such as vacancies and interstitiaIs which can result in dislocation loops and networks, voids and bubbles and causing changes in the layers in the thickness of the vessel. The microscopic defects produced in the vessel due to γ - radiation damage are referred to as radiation damage while the defects thus produced modify the macroscopic properties of the vessel which are also known as the radiation effects. These radiation damage effects are of major concern for materials used in nuclear energy production. In this study, the overall objective was to assess the effects of gamma radiation heating and damage in GHARR - I MNSR vessel by a well-developed Mathematical model, Analytical and Numerical solutions, simulating the radiation damage in the vessel. SRIM - TRIM Code was used as a computational tool to determine the displacement per atom (dpa) associated with radiation damage while implicit Control Volume Finite Difference Method was used to determine the temperature profile within the vessel due to γ - radiation heating respectively. The methodology adopted in assessing γ - radiation heating in the vessel involved development of the One-Dimensional Steady State Fourier Heat Conduction Equation with Volumetric Heat Generation both analytical and implicit Control Volume Finite Difference Method approach to determine the maximum temperature and

  11. MAXED, a computer code for the deconvolution of multisphere neutron spectrometer data using the maximum entropy method

    International Nuclear Information System (INIS)

    Reginatto, M.; Goldhagen, P.

    1998-06-01

    The problem of analyzing data from a multisphere neutron spectrometer to infer the energy spectrum of the incident neutrons is discussed. The main features of the code MAXED, a computer program developed to apply the maximum entropy principle to the deconvolution (unfolding) of multisphere neutron spectrometer data, are described, and the use of the code is illustrated with an example. A user's guide for the code MAXED is included in an appendix. The code is available from the authors upon request

  12. Comparison of a Label-Free Quantitative Proteomic Method Based on Peptide Ion Current Area to the Isotope Coded Affinity Tag Method

    Directory of Open Access Journals (Sweden)

    Young Ah Goo

    2008-01-01

    Full Text Available Recently, several research groups have published methods for the determination of proteomic expression profiling by mass spectrometry without the use of exogenously added stable isotopes or stable isotope dilution theory. These so-called label-free, methods have the advantage of allowing data on each sample to be acquired independently from all other samples to which they can later be compared in silico for the purpose of measuring changes in protein expression between various biological states. We developed label free software based on direct measurement of peptide ion current area (PICA and compared it to two other methods, a simpler label free method known as spectral counting and the isotope coded affinity tag (ICAT method. Data analysis by these methods of a standard mixture containing proteins of known, but varying, concentrations showed that they performed similarly with a mean squared error of 0.09. Additionally, complex bacterial protein mixtures spiked with known concentrations of standard proteins were analyzed using the PICA label-free method. These results indicated that the PICA method detected all levels of standard spiked proteins at the 90% confidence level in this complex biological sample. This finding confirms that label-free methods, based on direct measurement of the area under a single ion current trace, performed as well as the standard ICAT method. Given the fact that the label-free methods provide ease in experimental design well beyond pair-wise comparison, label-free methods such as our PICA method are well suited for proteomic expression profiling of large numbers of samples as is needed in clinical analysis.

  13. Elaboration of a computer code for the solution of a two-dimensional two-energy group diffusion problem using the matrix response method

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1980-12-01

    An analytical procedure to solve the neutron diffusion equation in two dimensions and two energy groups was developed. The response matrix method was used coupled with an expansion of the neutron flux in finite Fourier series. A computer code 'MRF2D' was elaborated to implement the above mentioned procedure for PWR reactor core calculations. Different core symmetry options are allowed by the code, which is also flexible enough to allow for improvements by means of algorithm optimization. The code performance was compared with a corner mesh finite difference code named TVEDIM by using a International Atomic Energy Agency (IAEA) standard problem. Computer processing time 12,7% smaller is required by the MRF2D code to reach the same precision on criticality eigenvalue. (Author) [pt

  14. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  15. Elaboration of a nodal method to solve the steady state multigroup diffusion equation. Study and use of the multigroup diffusion code DAHRA

    International Nuclear Information System (INIS)

    Halilou, A.; Lounici, A.

    1981-01-01

    The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method

  16. Novel methods for the molecular discrimination of Fasciola spp. on the basis of nuclear protein-coding genes.

    Science.gov (United States)

    Shoriki, Takuya; Ichikawa-Seki, Madoka; Suganuma, Keisuke; Naito, Ikunori; Hayashi, Kei; Nakao, Minoru; Aita, Junya; Mohanta, Uday Kumar; Inoue, Noboru; Murakami, Kenji; Itagaki, Tadashi

    2016-06-01

    Fasciolosis is an economically important disease of livestock caused by Fasciola hepatica, Fasciola gigantica, and aspermic Fasciola flukes. The aspermic Fasciola flukes have been discriminated morphologically from the two other species by the absence of sperm in their seminal vesicles. To date, the molecular discrimination of F. hepatica and F. gigantica has relied on the nucleotide sequences of the internal transcribed spacer 1 (ITS1) region. However, ITS1 genotypes of aspermic Fasciola flukes cannot be clearly differentiated from those of F. hepatica and F. gigantica. Therefore, more precise and robust methods are required to discriminate Fasciola spp. In this study, we developed PCR restriction fragment length polymorphism and multiplex PCR methods to discriminate F. hepatica, F. gigantica, and aspermic Fasciola flukes on the basis of the nuclear protein-coding genes, phosphoenolpyruvate carboxykinase and DNA polymerase delta, which are single locus genes in most eukaryotes. All aspermic Fasciola flukes used in this study had mixed fragment pattern of F. hepatica and F. gigantica for both of these genes, suggesting that the flukes are descended through hybridization between the two species. These molecular methods will facilitate the identification of F. hepatica, F. gigantica, and aspermic Fasciola flukes, and will also prove useful in etiological studies of fasciolosis. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  17. The impact of conventional dietary intake data coding methods on foods typically consumed by low-income African-American and White urban populations.

    Science.gov (United States)

    Mason, Marc A; Fanelli Kuczmarski, Marie; Allegro, Deanne; Zonderman, Alan B; Evans, Michele K

    2015-08-01

    Analysing dietary data to capture how individuals typically consume foods is dependent on the coding variables used. Individual foods consumed simultaneously, like coffee with milk, are given codes to identify these combinations. Our literature review revealed a lack of discussion about using combination codes in analysis. The present study identified foods consumed at mealtimes and by race when combination codes were or were not utilized. Duplicate analysis methods were performed on separate data sets. The original data set consisted of all foods reported; each food was coded as if it was consumed individually. The revised data set was derived from the original data set by first isolating coded foods consumed as individual items from those foods consumed simultaneously and assigning a code to designate a combination. Foods assigned a combination code, like pancakes with syrup, were aggregated and associated with a food group, defined by the major food component (i.e. pancakes), and then appended to the isolated coded foods. Healthy Aging in Neighborhoods of Diversity across the Life Span study. African-American and White adults with two dietary recalls (n 2177). Differences existed in lists of foods most frequently consumed by mealtime and race when comparing results based on original and revised data sets. African Americans reported consumption of sausage/luncheon meat and poultry, while ready-to-eat cereals and cakes/doughnuts/pastries were reported by Whites on recalls. Use of combination codes provided more accurate representation of how foods were consumed by populations. This information is beneficial when creating interventions and exploring diet-health relationships.

  18. A Bipartite Network-based Method for Prediction of Long Non-coding RNA–protein Interactions

    Directory of Open Access Journals (Sweden)

    Mengqu Ge

    2016-02-01

    Full Text Available As one large class of non-coding RNAs (ncRNAs, long ncRNAs (lncRNAs have gained considerable attention in recent years. Mutations and dysfunction of lncRNAs have been implicated in human disorders. Many lncRNAs exert their effects through interactions with the corresponding RNA-binding proteins. Several computational approaches have been developed, but only few are able to perform the prediction of these interactions from a network-based point of view. Here, we introduce a computational method named lncRNA–protein bipartite network inference (LPBNI. LPBNI aims to identify potential lncRNA–interacting proteins, by making full use of the known lncRNA–protein interactions. Leave-one-out cross validation (LOOCV test shows that LPBNI significantly outperforms other network-based methods, including random walk (RWR and protein-based collaborative filtering (ProCF. Furthermore, a case study was performed to demonstrate the performance of LPBNI using real data in predicting potential lncRNA–interacting proteins.

  19. Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo [Korea Advanced Institute of Science and Tehcnology, Daejeon (Korea, Republic of)

    2006-03-15

    There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis.

  20. Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

    International Nuclear Information System (INIS)

    Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo

    2006-03-01

    There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis

  1. Development of a computer code system for selecting off-site protective action in radiological accidents based on the multiobjective optimization method

    International Nuclear Information System (INIS)

    Ishigami, Tsutomu; Oyama, Kazuo

    1989-09-01

    This report presents a new method to support selection of off-site protective action in nuclear reactor accidents, and provides a user's manual of a computer code system, PRASMA, developed using the method. The PRASMA code system gives several candidates of protective action zones of evacuation, sheltering and no action based on the multiobjective optimization method, which requires objective functions and decision variables. We have assigned population risks of fatality, injury and cost as the objective functions, and distance from a nuclear power plant characterizing the above three protective action zones as the decision variables. (author)

  2. Dynamic Server-Based KML Code Generator Method for Level-of-Detail Traversal of Geospatial Data

    Science.gov (United States)

    Baxes, Gregory; Mixon, Brian; Linger, TIm

    2013-01-01

    Web-based geospatial client applications such as Google Earth and NASA World Wind must listen to data requests, access appropriate stored data, and compile a data response to the requesting client application. This process occurs repeatedly to support multiple client requests and application instances. Newer Web-based geospatial clients also provide user-interactive functionality that is dependent on fast and efficient server responses. With massively large datasets, server-client interaction can become severely impeded because the server must determine the best way to assemble data to meet the client applications request. In client applications such as Google Earth, the user interactively wanders through the data using visually guided panning and zooming actions. With these actions, the client application is continually issuing data requests to the server without knowledge of the server s data structure or extraction/assembly paradigm. A method for efficiently controlling the networked access of a Web-based geospatial browser to server-based datasets in particular, massively sized datasets has been developed. The method specifically uses the Keyhole Markup Language (KML), an Open Geospatial Consortium (OGS) standard used by Google Earth and other KML-compliant geospatial client applications. The innovation is based on establishing a dynamic cascading KML strategy that is initiated by a KML launch file provided by a data server host to a Google Earth or similar KMLcompliant geospatial client application user. Upon execution, the launch KML code issues a request for image data covering an initial geographic region. The server responds with the requested data along with subsequent dynamically generated KML code that directs the client application to make follow-on requests for higher level of detail (LOD) imagery to replace the initial imagery as the user navigates into the dataset. The approach provides an efficient data traversal path and mechanism that can be

  3. Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

    1988-01-01

    This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)

  4. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

  5. About the use of the Monte-Carlo code based tracing algorithm and the volume fraction method for S n full core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Gurevich, M. I.; Oleynik, D. S. [RRC Kurchatov Inst., Kurchatov Sq., 1, 123182, Moscow (Russian Federation); Russkov, A. A.; Voloschenko, A. M. [Keldysh Inst. of Applied Mathematics, Miusskaya Sq., 4, 125047, Moscow (Russian Federation)

    2006-07-01

    The tracing algorithm that is implemented in the geometrical module of Monte-Carlo transport code MCU is applied to calculate the volume fractions of original materials by spatial cells of the mesh that overlays problem geometry. In this way the 3D combinatorial geometry presentation of the problem geometry, used by MCU code, is transformed to the user defined 2D or 3D bit-mapped ones. Next, these data are used in the volume fraction (VF) method to approximate problem geometry by introducing additional mixtures for spatial cells, where a few original materials are included. We have found that in solving realistic 2D and 3D core problems a sufficiently fast convergence of the VF method takes place if the spatial mesh is refined. Virtually, the proposed variant of implementation of the VF method seems as a suitable geometry interface between Monte-Carlo and S{sub n} transport codes. (authors)

  6. Development of a neutron transport code many-group two-dimensional heterogeneous calculations by the method of characteristics

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2000-01-01

    The method of characteristics (MOC) is gaining increased popularity in the reactor physics community all over the world because it gives a new degree of freedom in nuclear reactor analysis. The MARIKO code solves the neutron transport equation by the MOC in two-dimensional real geometry. The domain of solution can be a rectangle or right hexagon with periodic boundary conditions on the outer boundary. Any reasonable symmetry inside the domain can be fully accounted for. The geometry is described in three levels-macro-cells, cells, and regions. The macro-cells and cells can be any polygon. The outer boundary of a region can be any combination of straight line and circular arc segments. Any level of embedded regions is allowed. Procedures for automatic geometry description of hexagonal fuel assemblies and reflector macro-cells have been developed. The initial ray tracing procedure is performed for the full rectangular or hexagonal domain, but only azimuthal angles in the smallest symmetry interval are tracked. (Authors)

  7. Calculation of accurate albedo boundary conditions for three-dimensional nodal diffusion codes by the method of characteristics

    International Nuclear Information System (INIS)

    Petkov, Petko T.

    2000-01-01

    Most of the few-group three-dimensional nodal diffusion codes used for neutronics calculations of the WWER reactors use albedo type boundary conditions on the core-reflector boundary. The conventional albedo are group-to-group reflection probabilities, defined on each outer node face. The method of characteristics is used to calculate accurate albedo by the following procedure. A many-group two-dimensional heterogeneous core-reflector problem, including a sufficient part of the core and detailed description of the adjacent reflector, is solved first. From this solution the angular flux on the core-reflector boundary is calculated in all groups for all traced neutron directions. Accurate boundary conditions can be calculated for the radial, top and bottom reflectors as well as for the absorber part of the WWER-440 reactor control assemblies. The algorithm can be used to estimate also albedo, coupling outer node faces on the radial reflector in the axial direction. Numerical results for the WWER-440 reactor are presented. (Authors)

  8. Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

    International Nuclear Information System (INIS)

    Cai, Li

    2014-01-01

    In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

  9. A Dual-Channel Acquisition Method Based on Extended Replica Folding Algorithm for Long Pseudo-Noise Code in Inter-Satellite Links.

    Science.gov (United States)

    Zhao, Hongbo; Chen, Yuying; Feng, Wenquan; Zhuang, Chen

    2018-05-25

    Inter-satellite links are an important component of the new generation of satellite navigation systems, characterized by low signal-to-noise ratio (SNR), complex electromagnetic interference and the short time slot of each satellite, which brings difficulties to the acquisition stage. The inter-satellite link in both Global Positioning System (GPS) and BeiDou Navigation Satellite System (BDS) adopt the long code spread spectrum system. However, long code acquisition is a difficult and time-consuming task due to the long code period. Traditional folding methods such as extended replica folding acquisition search technique (XFAST) and direct average are largely restricted because of code Doppler and additional SNR loss caused by replica folding. The dual folding method (DF-XFAST) and dual-channel method have been proposed to achieve long code acquisition in low SNR and high dynamic situations, respectively, but the former is easily affected by code Doppler and the latter is not fast enough. Considering the environment of inter-satellite links and the problems of existing algorithms, this paper proposes a new long code acquisition algorithm named dual-channel acquisition method based on the extended replica folding algorithm (DC-XFAST). This method employs dual channels for verification. Each channel contains an incoming signal block. Local code samples are folded and zero-padded to the length of the incoming signal block. After a circular FFT operation, the correlation results contain two peaks of the same magnitude and specified relative position. The detection process is eased through finding the two largest values. The verification takes all the full and partial peaks into account. Numerical results reveal that the DC-XFAST method can improve acquisition performance while acquisition speed is guaranteed. The method has a significantly higher acquisition probability than folding methods XFAST and DF-XFAST. Moreover, with the advantage of higher detection

  10. A Dual-Channel Acquisition Method Based on Extended Replica Folding Algorithm for Long Pseudo-Noise Code in Inter-Satellite Links

    Directory of Open Access Journals (Sweden)

    Hongbo Zhao

    2018-05-01

    Full Text Available Inter-satellite links are an important component of the new generation of satellite navigation systems, characterized by low signal-to-noise ratio (SNR, complex electromagnetic interference and the short time slot of each satellite, which brings difficulties to the acquisition stage. The inter-satellite link in both Global Positioning System (GPS and BeiDou Navigation Satellite System (BDS adopt the long code spread spectrum system. However, long code acquisition is a difficult and time-consuming task due to the long code period. Traditional folding methods such as extended replica folding acquisition search technique (XFAST and direct average are largely restricted because of code Doppler and additional SNR loss caused by replica folding. The dual folding method (DF-XFAST and dual-channel method have been proposed to achieve long code acquisition in low SNR and high dynamic situations, respectively, but the former is easily affected by code Doppler and the latter is not fast enough. Considering the environment of inter-satellite links and the problems of existing algorithms, this paper proposes a new long code acquisition algorithm named dual-channel acquisition method based on the extended replica folding algorithm (DC-XFAST. This method employs dual channels for verification. Each channel contains an incoming signal block. Local code samples are folded and zero-padded to the length of the incoming signal block. After a circular FFT operation, the correlation results contain two peaks of the same magnitude and specified relative position. The detection process is eased through finding the two largest values. The verification takes all the full and partial peaks into account. Numerical results reveal that the DC-XFAST method can improve acquisition performance while acquisition speed is guaranteed. The method has a significantly higher acquisition probability than folding methods XFAST and DF-XFAST. Moreover, with the advantage of higher

  11. Criticality calculations on pebble-bed HTR-PROTEUS configuration as a validation for the pseudo-scattering tracking method implemented in the MORET 5 Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Forestier, Benoit; Miss, Joachim; Bernard, Franck; Dorval, Aurelien [Institut de Radioprotection et Surete Nucleaire, Fontenay aux Roses (France); Jacquet, Olivier [Independent consultant (France); Verboomen, Bernard [Belgian Nuclear Research Center - SCK-CEN (Belgium)

    2008-07-01

    The MORET code is a three dimensional Monte Carlo criticality code. It is designed to calculate the effective multiplication factor (k{sub eff}) of any geometrical configuration as well as the reaction rates in the various volumes and the neutron leakage out of the system. A recent development for the MORET code consists of the implementation of an alternate neutron tracking method, known as the pseudo-scattering tracking method. This method has been successfully implemented in the MORET code and its performances have been tested by mean of an extensive parametric study on very simple geometrical configurations. In this context, the goal of the present work is to validate the pseudo-scattering method against realistic configurations. In this perspective, pebble-bed cores are particularly well-adapted cases to model, as they exhibit large amount of volumes stochastically arranged on two different levels (the pebbles in the core and the TRISO particles inside each pebble). This paper will introduce the techniques and methods used to model pebble-bed cores in a realistic way. The results of the criticality calculations, as well as the pseudo-scattering tracking method performance in terms of computation time, will also be presented. (authors)

  12. Japanese standard method for safety evaluation using best estimate code based on uncertainty and scaling analyses with statistical approach

    International Nuclear Information System (INIS)

    Mizokami, Shinya; Hotta, Akitoshi; Kudo, Yoshiro; Yonehara, Tadashi; Watada, Masayuki; Sakaba, Hiroshi

    2009-01-01

    Current licensing practice in Japan consists of using conservative boundary and initial conditions(BIC), assumptions and analytical codes. The safety analyses for licensing purpose are inherently deterministic. Therefore, conservative BIC and assumptions, such as single failure, must be employed for the analyses. However, using conservative analytical codes are not considered essential. The standard committee of Atomic Energy Society of Japan(AESJ) has drawn up the standard for using best estimate codes for safety analyses in 2008 after three-years of discussions reflecting domestic and international recent findings. (author)

  13. Narrowing of the middle cerebral artery: artificial intelligence methods and comparison of transcranial color coded duplex sonography with conventional TCD.

    Science.gov (United States)

    Swiercz, Miroslaw; Swiat, Maciej; Pawlak, Mikolaj; Weigele, John; Tarasewicz, Roman; Sobolewski, Andrzej; Hurst, Robert W; Mariak, Zenon D; Melhem, Elias R; Krejza, Jaroslaw

    2010-01-01

    The goal of the study was to compare performances of transcranial color-coded duplex sonography (TCCS) and transcranial Doppler sonography (TCD) in the diagnosis of the middle cerebral artery (MCA) narrowing in the same population of patients using statistical and nonstatistical intelligent models for data analysis. We prospectively collected data from 179 consecutive routine digital subtraction angiography (DSA) procedures performed in 111 patients (mean age 54.17+/-14.4 years; 59 women, 52 men) who underwent TCD and TCCS examinations simultaneously. Each patient was examined independently using both ultrasound techniques, 267 M1 segments of MCA were assessed and narrowings were classified as 50% lumen reduction. Diagnostic performance was estimated by two statistical and two artificial neural networks (ANN) classification methods. Separate models were constructed for the TCD and TCCS sonographic data, as well as for detection of "any narrowing" and "severe narrowing" of the MCA. Input for each classifier consisted of the peak-systolic, mean and end-diastolic velocities measured with each sonographic method; the output was MCA narrowing. Arterial narrowings less or equal 50% of lumen reduction were found in 55 and >50% narrowings in 26 out of 267 arteries, as indicated by DSA. In the category of "any narrowing" the rate of correct assignment by all models was 82% to 83% for TCCS and 79% to 81% for TCD. In the diagnosis of >50% narrowing the overall classification accuracy remained in the range of 89% to 90% for TCCS data and 90% to 91% for TCD data. For the diagnosis of any narrowing, the sensitivity of the TCCS was significantly higher than that of the TCD, while for diagnosis of >50% MCA narrowing, sensitivity of the TCCS was similar to sensitivity of the TCD. Our study showed that TCCS outperforms conventional TCD in detection of diagnosis of >50% MCA narrowing. (E-mail: jaroslaw.krejza@uphs.upenn.edu).

  14. Principle of the determination of neutron multiplication coefficients by the Monte Carlo method. Application. Description of a code for ibm 360-75; Principe de la determination des coefficients de multiplication neutronique par methode de Monte-Carlo. Application. Description d'un code pour IBM 360-75

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, J; Parisot, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The determination of neutron multiplication coefficients by the Monte Carlo method can be carried out in different ways; the are first examined particularly complex geometries; it makes use of multi-group isotropic cross sections. The performances of this code are illustrated by some examples. (author) [French] La determination des coefficients de multiplication neutronique par methode de Monte Carlo peut se faire par differentes voies, elles sont successivement examinees et comparees. On en deduit un code rapide pour des geometries particulierement complexes, il utilise des sections efficaces multigroupes isotropes. Les performances de ce code sont demontrees par quelques exemples. (auteur)

  15. The materiality of Code

    DEFF Research Database (Denmark)

    Soon, Winnie

    2014-01-01

    This essay studies the source code of an artwork from a software studies perspective. By examining code that come close to the approach of critical code studies (Marino, 2006), I trace the network artwork, Pupufu (Lin, 2009) to understand various real-time approaches to social media platforms (MSN......, Twitter and Facebook). The focus is not to investigate the functionalities and efficiencies of the code, but to study and interpret the program level of code in order to trace the use of various technological methods such as third-party libraries and platforms’ interfaces. These are important...... to understand the socio-technical side of a changing network environment. Through the study of code, including but not limited to source code, technical specifications and other materials in relation to the artwork production, I would like to explore the materiality of code that goes beyond technical...

  16. MVP/GMVP 2: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

    2005-06-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)

  17. A criticality safety analysis code using a vectorized Monte Carlo method on the HITAC S-810 supercomputer

    International Nuclear Information System (INIS)

    Morimoto, Y.; Maruyama, H.

    1987-01-01

    A vectorized Monte Carlo criticality safety analysis code has been developed on the vector supercomputer HITAC S-810. In this code, a multi-particle tracking algorithm was adopted for effective utilization of the vector processor. A flight analysis with pseudo-scattering was developed to reduce the computational time needed for flight analysis, which represents the bulk of computational time. This new algorithm realized a speed-up of factor 1.5 over the conventional flight analysis. The code also adopted the multigroup cross section constants library of the Bodarenko type with 190 groups, with 132 groups being for fast and epithermal regions and 58 groups being for the thermal region. Evaluation work showed that this code reproduce the experimental results to an accuracy of about 1 % for the effective neutron multiplication factor. (author)

  18. Assessment of 12 CHF prediction methods, for an axially non-uniform heat flux distribution, with the RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Ferrouk, M. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria)], E-mail: m_ferrouk@yahoo.fr; Aissani, S. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria); D' Auria, F.; DelNevo, A.; Salah, A. Bousbia [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (Italy)

    2008-10-15

    The present article covers the evaluation of the performance of twelve critical heat flux methods/correlations published in the open literature. The study concerns the simulation of an axially non-uniform heat flux distribution with the RELAP5 computer code in a single boiling water reactor channel benchmark problem. The nodalization scheme employed for the considered particular geometry, as modelled in RELAP5 code, is described. For this purpose a review of critical heat flux models/correlations applicable to non-uniform axial heat profile is provided. Simulation results using the RELAP5 code and those obtained from our computer program, based on three type predictions methods such as local conditions, F-factor and boiling length average approaches were compared.

  19. Development of next generation code system as an engineering modeling language (5). Investigation on restructuring method of conventional codes into two-layer system

    International Nuclear Information System (INIS)

    Yokoyama, Kenji

    2006-10-01

    A proposed method for gradually restructuring to the two-level system of next generation analysis system by reusing the conventional analysis system, called 'incremental method', was applied and evaluated. The following functions were selected for the evaluation: Neutron diffusion calculation for the three-dimensional XYZ system based on finite differential method, and input utilities of the cross-section data file used in the conventional system. In order to evaluate the effect of the restructuring, 'Module Coupling Index (MCI)' and 'McCabe's Cyclomatic Complexity (MCC)' were used for quantifying the quality of the modular design and the complexity of the program sequence of each module. Although MCIs of each module before restructuring were mainly 6 - 7 degrees, it was possible to reduce them to under 4 degrees in most module by restructuring with the incremental method. And, it is found that the modules under 4 degrees of MCI can be easily combined with different programming languages, which are necessary for building the two-layer system. In the meantime, MCCs in most module before restructuring were over 20 and some were over 50. The incremental method could reduce them to under 10 when C++ was used, and reduce them to under 20 when FORTRAN was used. It is correspondent to reduction of the error frequency occurred in its modification from 20 - 40% to 5 - 10%. The total number of MCC could be reduced to 1/3 when C++ was used, and to 1/2 when FORTRAN was used. By using the restructured functions in the present study and some previously developed functions, a reactor analysis tool was systematized and applied to criticality analysis of the Experimental Fast Reactor 'JOYO' MR-I. In addition, the following two functionality expansion tests were performed: To add cross section direct perturbation functionality, and to add control rod criticality position search functionality. In the tests, both the functionality expansions were carried out satisfying the condition

  20. Performance of four computer-coded verbal autopsy methods for cause of death assignment compared with physician coding on 24,000 deaths in low- and middle-income countries

    Science.gov (United States)

    2014-01-01

    Background Physician-coded verbal autopsy (PCVA) is the most widely used method to determine causes of death (CODs) in countries where medical certification of death is uncommon. Computer-coded verbal autopsy (CCVA) methods have been proposed as a faster and cheaper alternative to PCVA, though they have not been widely compared to PCVA or to each other. Methods We compared the performance of open-source random forest, open-source tariff method, InterVA-4, and the King-Lu method to PCVA on five datasets comprising over 24,000 verbal autopsies from low- and middle-income countries. Metrics to assess performance were positive predictive value and partial chance-corrected concordance at the individual level, and cause-specific mortality fraction accuracy and cause-specific mortality fraction error at the population level. Results The positive predictive value for the most probable COD predicted by the four CCVA methods averaged about 43% to 44% across the datasets. The average positive predictive value improved for the top three most probable CODs, with greater improvements for open-source random forest (69%) and open-source tariff method (68%) than for InterVA-4 (62%). The average partial chance-corrected concordance for the most probable COD predicted by the open-source random forest, open-source tariff method and InterVA-4 were 41%, 40% and 41%, respectively, with better results for the top three most probable CODs. Performance generally improved with larger datasets. At the population level, the King-Lu method had the highest average cause-specific mortality fraction accuracy across all five datasets (91%), followed by InterVA-4 (72% across three datasets), open-source random forest (71%) and open-source tariff method (54%). Conclusions On an individual level, no single method was able to replicate the physician assignment of COD more than about half the time. At the population level, the King-Lu method was the best method to estimate cause-specific mortality

  1. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th.; Nimal, J.C.; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  2. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th; Nimal, J C; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  3. Continuous-energy adjoint flux and perturbation calculation using the iterated fission probability method in Monte-Carlo code TRIPOLI-4 and underlying applications

    International Nuclear Information System (INIS)

    Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.

    2013-01-01

    The first goal of this paper is to present an exact method able to precisely evaluate very small reactivity effects with a Monte Carlo code (<10 pcm). it has been decided to implement the exact perturbation theory in TRIPOLI-4 and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4 is described. To illustrate the efficiency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the 'direct' estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the 'direct' method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. It offers the possibility to split reactivity contributions on both isotopes and reactions. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters

  4. The Monte Carlo method for shielding calculations analysis by MORSE code of a streaming case in the CAORSO BWR power reactor shielding (Italy)

    International Nuclear Information System (INIS)

    Zitouni, Y.

    1987-04-01

    In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us

  5. Structural evaluation method for class 1 vessels by using elastic-plastic finite element analysis in code case of JSME rules on design and construction

    International Nuclear Information System (INIS)

    Asada, Seiji; Hirano, Takashi; Nagata, Tetsuya; Kasahara, Naoto

    2008-01-01

    A structural evaluation method by using elastic-plastic finite element analysis has been developed and published as a code case of Rules on Design and Construction for Nuclear Power Plants (The First Part: Light Water Reactor Structural Design Standard) in the JSME Codes for Nuclear Power Generation Facilities. Its title is 'Alternative Structural Evaluation Criteria for Class 1 Vessels Based on Elastic-Plastic Finite Element Analysis' (NC-CC-005). This code case applies elastic-plastic analysis to evaluation of such failure modes as plastic collapse, thermal ratchet, fatigue and so on. Advantage of this evaluation method is free from stress classification, consistently use of Mises stress and applicability to complex 3-dimensional structures which are hard to be treated by the conventional stress classification method. The evaluation method for plastic collapse has such variation as the Lower Bound Approach Method, Twice-Elastic-Slope Method and Elastic Compensation Method. Cyclic Yield Area (CYA) based on elastic analysis is applied to screening evaluation of thermal ratchet instead of secondary stress evaluation, and elastic-plastic analysis is performed when the CYA screening criteria is not satisfied. Strain concentration factors can be directly calculated based on elastic-plastic analysis. (author)

  6. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  7. Application of low bitrate image coding to surveillance of electric power facilities. Part 1. Proposal of low bitrate coding for surveillance of electric power facilities and examination of facilities region extraction method; Denryoku setsubi kanshi eno tei rate fugoka hoshiki no tekiyo. 1. Setsubi kanshiyo fugoka hoshiki no teian to setsubi ryoiki chushutsuho no kento

    Energy Technology Data Exchange (ETDEWEB)

    Murata, H.; Ishino, R. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    1996-03-01

    Current status of low bitrate image coding has been investigated, and a low bitrate coding suitable for the surveillance of electric power facilities has been proposed, to extract its problems to be solved. For the conventional image coding, the waveform coding has been used by which the images are processed as signals. While, for the MPEG-4, a coding method with considering the image information has been proposed. For these coding methods, however, image information lacks details primarily, when lowering the bitrate. Accordingly, these methods can not be applied when the details in the images are important, such as in the case of surveillance of facilities. Then, the coding method has been proposed by expanding the partially detailed coding, and by separating constituent images of facilities, such as power cables and steel towers, designated by operators. It is the special feature of this method that the method can easily respond to the low bitrate and the detailed information can be conserved by using the structure extraction coding for the designated partial image which is generally processed by the low bitrate waveform coding. 29 refs., 17 figs., 1 tab.

  8. Analysis Code - Data Analysis in 'Leveraging Multiple Statistical Methods for Inverse Prediction in Nuclear Forensics Applications' (LMSMIPNFA) v. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    2018-03-19

    R code that performs the analysis of a data set presented in the paper ‘Leveraging Multiple Statistical Methods for Inverse Prediction in Nuclear Forensics Applications’ by Lewis, J., Zhang, A., Anderson-Cook, C. It provides functions for doing inverse predictions in this setting using several different statistical methods. The data set is a publicly available data set from a historical Plutonium production experiment.

  9. Modification of the fast fourier transform-based method by signal mirroring for accuracy quantification of thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Ha, Tae Wook; Jeong, Jae Jun; Choi, Ki Yong

    2017-01-01

    A thermal–hydraulic system code is an essential tool for the design and safety analysis of a nuclear power plant, and its accuracy quantification is very important for the code assessment and applications. The fast Fourier transform-based method (FFTBM) by signal mirroring (FFTBM-SM) has been used to quantify the accuracy of a system code by using a comparison of the experimental data and the calculated results. The method is an improved version of the FFTBM, and it is known that the FFTBM-SM judges the code accuracy in a more consistent and unbiased way. However, in some applications, unrealistic results have been obtained. In this study, it was found that accuracy quantification by FFTBM-SM is dependent on the frequency spectrum of the fast Fourier transform of experimental and error signals. The primary objective of this study is to reduce the frequency dependency of FFTBM-SM evaluation. For this, it was proposed to reduce the cut off frequency, which was introduced to cut off spurious contributions, in FFTBM-SM. A method to determine an appropriate cut off frequency was also proposed. The FFTBM-SM with the modified cut off frequency showed a significant improvement of the accuracy quantification

  10. MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  11. Modification of the fast fourier transform-based method by signal mirroring for accuracy quantification of thermal-hydraulic system code

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Tae Wook; Jeong, Jae Jun [School of Mechanical Engineering, Pusan National University, Busan (Korea, Republic of); Choi, Ki Yong [Korea Atomic Energy Research Institute (KAERI), Daejeon (Korea, Republic of)

    2017-08-15

    A thermal–hydraulic system code is an essential tool for the design and safety analysis of a nuclear power plant, and its accuracy quantification is very important for the code assessment and applications. The fast Fourier transform-based method (FFTBM) by signal mirroring (FFTBM-SM) has been used to quantify the accuracy of a system code by using a comparison of the experimental data and the calculated results. The method is an improved version of the FFTBM, and it is known that the FFTBM-SM judges the code accuracy in a more consistent and unbiased way. However, in some applications, unrealistic results have been obtained. In this study, it was found that accuracy quantification by FFTBM-SM is dependent on the frequency spectrum of the fast Fourier transform of experimental and error signals. The primary objective of this study is to reduce the frequency dependency of FFTBM-SM evaluation. For this, it was proposed to reduce the cut off frequency, which was introduced to cut off spurious contributions, in FFTBM-SM. A method to determine an appropriate cut off frequency was also proposed. The FFTBM-SM with the modified cut off frequency showed a significant improvement of the accuracy quantification.

  12. MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  13. Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1983-06-01

    The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

  14. Comparison of Monte Carlo simulations of photon/electron dosimetry in microscale applications

    International Nuclear Information System (INIS)

    Poneja, O.P.; Chawla, R.; Negreanu, C.; Stepanek, J.

    2003-01-01

    It is important to establish reliable calculational tools to plan and analyse representative microdosimetry experiments in the context of microbeam radiation therapy development. In this paper, an attempt has been made to investigate the suitability of the MCNP4C Monte Carlo code to adequately model photon/electron transport over micron distances. The case of a single cylindrical microbeam of 25-micron diameter incident on a water phantom has been simulated in detail with both MCNP4C and the code PSI-GEANT, for different incident photon energies, to get absorbed dose distributions at various depths, with and without electron transport being considered. In addition, dose distributions calculated for a single microbeam with a photon spectrum representative of the European Synchrotron Radiation Facility (ESRF) have been compared. Finally, a large number of cylindrical microbeams ( a total of 2601 beams, placed on a 200-micron square pitch, covering an area of lcm 2 ) incident on a water phantom have been considered to study cumulative radial dose distributions at different depths. From these distributions, ratios of peak (within the microbeam) to valley (mid-point along the diagonal connecting two microbeams) dose values have been determined. The various comparisons with PSI-GEANT results have shown that MCNP4C, with its high flexibility in terms of its numerous source and geometry description options, variance reduction methods, detailed error analysis, statistical checks and different tally types, can be a valuable tool for the analysis of microbeam experiments. Copyright (2003) Australasian College of Physical Scientists and Engineers in Medicine

  15. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  16. Life With and Without Coding: Two Methods for Early-Stage Data Analysis in Qualitative Research Aiming at Causal Explanations

    NARCIS (Netherlands)

    Gläser, Jochen; Laudel, Grit

    2013-01-01

    Qualitative research aimed at "mechanismic" explanations poses specific challenges to qualitative data analysis because it must integrate existing theory with patterns identified in the data. We explore the utilization of two methods—coding and qualitative content analysis—for the first steps in the

  17. Surviving "Payment by Results": a simple method of improving clinical coding in burn specialised services in the United Kingdom.

    Science.gov (United States)

    Wallis, Katy L; Malic, Claudia C; Littlewood, Sonia L; Judkins, Keith; Phipps, Alan R

    2009-03-01

    Coding inpatient episodes plays an important role in determining the financial remuneration of a clinical service. Insufficient or incomplete data may have very significant consequences on its viability. We created a document that improves the coding process in our Burns Centre. At Yorkshire Regional Burns Centre an inpatient summary sheet was designed to prospectively record and present essential information on a daily basis, for use in the coding process. The level of care was also recorded. A 3-month audit was conducted to assess the efficacy of the new forms. Forty-nine patients were admitted to the Burns Centre with a mean age of 27.6 years and TBSA ranging from 0.5% to 65%. The total stay in the Burns Centre was 758 days, of which 22% were at level B3-B5 and 39% at level B2. The use of the new discharge document identified potential income of about 500,000 GB pound sterling at our local daily tariffs for high dependency and intensive care. The new form is able to ensure a high quality of coding with a possible direct impact on the financial resources accrued for burn care.

  18. A two-phase pressure drop calculation code based on a new method with a correlation factor obtained from an assessment of existing correlations

    International Nuclear Information System (INIS)

    Chun, Moon Hyun; Oh, Jae Guen

    1989-01-01

    Ten methods of the total two-phase pressure drop prediction based on five existing models and correlations have been examined for their accuracy and applicability to pressurized water reactor conditions. These methods were tested against 209 experimental data of local and bulk boiling conditions: Each correlations were evaluated for different ranges of pressure, mass velocity and quality, and best performing models were identified for each data subsets. A computer code entitled 'K-TWOPD' has been developed to calculate the total two phase pressure drop using the best performing existing correlations for a specific property range and a correction factor to compensate for the predicted error of the selected correlations. Assessment of this code shows that the present method fits all the available data within ±11% at a 95% confidence level compared with ± 25% for the existing correlations. (Author)

  19. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  20. Adapting the coping in deliberation (CODE) framework: a multi-method approach in the context of familial ovarian cancer risk management.

    Science.gov (United States)

    Witt, Jana; Elwyn, Glyn; Wood, Fiona; Rogers, Mark T; Menon, Usha; Brain, Kate

    2014-11-01

    To test whether the coping in deliberation (CODE) framework can be adapted to a specific preference-sensitive medical decision: risk-reducing bilateral salpingo-oophorectomy (RRSO) in women at increased risk of ovarian cancer. We performed a systematic literature search to identify issues important to women during deliberations about RRSO. Three focus groups with patients (most were pre-menopausal and untested for genetic mutations) and 11 interviews with health professionals were conducted to determine which issues mattered in the UK context. Data were used to adapt the generic CODE framework. The literature search yielded 49 relevant studies, which highlighted various issues and coping options important during deliberations, including mutation status, risks of surgery, family obligations, physician recommendation, peer support and reliable information sources. Consultations with UK stakeholders confirmed most of these factors as pertinent influences on deliberations. Questions in the generic framework were adapted to reflect the issues and coping options identified. The generic CODE framework was readily adapted to a specific preference-sensitive medical decision, showing that deliberations and coping are linked during deliberations about RRSO. Adapted versions of the CODE framework may be used to develop tailored decision support methods and materials in order to improve patient-centred care. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  1. Fire risk assessment method under the technical building code; Metodo de evaluacion del reisgo de incendio en el marco del Codigo Tecnico de la Edificacion

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Martin, J. C.; Diaz-Diaz, R.; Santos-Garcia, R.

    2010-07-01

    The high complexity of measuring and comparing the risk level of performance based design allowed by the Spain Building Code (CTE) and the lack of Spanish legislation on the matter makes a fire risk evaluation tool highly useful guiding the project's and the control authority over the feasibility of these complicated projects. A fire Risk Evaluation Method within the frame of the Spanish Building Code (MEREDICTE) has been developed trying to balance input sources, simplicity and clearness in its use. The number of parameters implied guarantees sound, exhaustive and reliable outputs. MEREDICTE is made of 73 parameters: 31 calculate potential risk and 42 obtain protection level. In an orientative way, MEREDICTE's parameters triple those of the most referred and extended evaluation method: the Gretener Method. The article shows the MEREDICTE technique and foundations, the methodology used in its investigation and development, its most significant innovation and its possible applications. Regarding the protection level, its formulation and applications re referred to the Windsor buildings of Madrid comparing the results obtained on the building as it was devastated by the fire of February 2005 and whether it fulfilled the Spanish building code conditions. (Author) 8 refs.

  2. Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR

    International Nuclear Information System (INIS)

    Chaudri, Khurrum Saleem; Su Yali; Chen Ronghua; Tian Wenxi; Su Guanghui; Qiu Suizheng

    2012-01-01

    Highlights: ► A tool is developed for coupled neutronics/thermal-hydraulic analysis for SCWR. ► For thermal hydraulic analysis, a sub-channel code SACoS is developed and verified. ► Coupled analysis agree quite well with the reference calculations. ► Different choice of important parameters makes huge difference in design calculations. - Abstract: Supercritical Water Reactor (SCWR) is one of the promising reactors from the list of fourth generation of nuclear reactors. High thermal efficiency and low cost of electricity make it an attractive option in the era of growing energy demand. An almost seven fold density variation for coolant/moderator along the active height does not allow the use of constant density assumption for design calculations, as used for previous generations of reactors. The advancement in computer technology gives us the superior option of performing coupled analysis. Thermal hydraulics calculations of supercritical water systems present extra challenges as not many computational tools are available to perform that job. This paper introduces a new sub-channel code called Sub-channel Analysis Code of SCWR (SACoS) and its application in coupled analyses of High Performance Light Water Reactor (HPLWR). SACoS can compute the basic thermal hydraulic parameters needed for design studies of a supercritical water reactor. Multiple heat transfer and pressure drop correlations are incorporated in the code according to the flow regime. It has the additional capability of calculating the thermal hydraulic parameters of moderator flowing in water box and between fuel assemblies under co-current or counter current flow conditions. Using MCNP4c and SACoS, a coupled system has been developed for SCWR design analyses. The developed coupled system is verified by performing and comparing HPLWR calculations. The results were found to be in very good agreement. Significant difference between the results was seen when Doppler feedback effect was included in

  3. Evaluation Codes from an Affine Veriety Code Perspective

    DEFF Research Database (Denmark)

    Geil, Hans Olav

    2008-01-01

    Evaluation codes (also called order domain codes) are traditionally introduced as generalized one-point geometric Goppa codes. In the present paper we will give a new point of view on evaluation codes by introducing them instead as particular nice examples of affine variety codes. Our study...... includes a reformulation of the usual methods to estimate the minimum distances of evaluation codes into the setting of affine variety codes. Finally we describe the connection to the theory of one-pointgeometric Goppa codes. Contents 4.1 Introduction...... . . . . . . . . . . . . . . . . . . . . . . . 171 4.9 Codes form order domains . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 4.10 One-point geometric Goppa codes . . . . . . . . . . . . . . . . . . . . . . . . 176 4.11 Bibliographical Notes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 References...

  4. Dosimetric analysis of radiation sources for use dermatological lesions

    International Nuclear Information System (INIS)

    Tada, Ariane

    2010-01-01

    Skin lesions undergoing therapy with radiation sources may have different patterns of malignancy. Malignant lesions or cancer most commonly found in radiotherapy services are carcinomas. Radiation therapy in skin lesions is performed with low penetration beams and orthovoltage X-rays, electron beams and radioactive sources ( 192 Ir, 198 Au, e 90 Sr) arranged on a surface mold or in metal applicator. This study aims to analyze the therapeutic radiation dose profile produced by radiation sources used in skin lesions radiotherapy procedures . Experimental measurements for the analysis of dosimetric radiation sources were compared with calculations obtained from a computer system based on the Monte Carlo Method. Computational results had a good agreement with the experimental measurements. Experimental measurements and computational results by the MCNP4C code were both physically consistent as expected. These experimental measurements compared with calculations using the MCNP-4C code have been used to validate the calculations obtained by MCNP code and to provide a reliable medical application for each clinical case. (author)

  5. Experimental determination of spectral ratios and of neutrons energy spectrum in the fuel of the IPEN/MB-01 nuclear reactor

    International Nuclear Information System (INIS)

    Nunes, Beatriz Guimaraes

    2012-01-01

    This study aims to determine the spectral ratios and the neutron energy spectrum inside the fuel of IPEN/MB-01 Nuclear Reactor. These parameters are of great importance to accurately determine spectral physical parameters of nuclear reactors like reaction rates, fuel lifetime and also security parameters such as reactivity. For the experiment, activation detectors in the form of thin metal foils were introduced in a collapsible fuel rod. Then the rod was placed in the central position of the core which has a standard rectangular configuration of 26 x 28 fuel rods. There were used activation detectors from different elements such Au-197, U-238, Sc-45, Ni-58, Mg-24, Ti-47 and In-115 to cover a large range of the neutron energy spectrum. After the irradiation, the activation detectors were submitted to gamma spectrometry using a counting system with high purity Germanium, to obtain the reaction rates (saturation activity) per target nucleus. The spectral ratios were compared with calculated values obtained by the Monte Carlo method using the MCNP-4C code. The neutron energy spectrum was obtained inside the fuel rod using the SANDBP code with an input spectrum obtained by the MCNP-4C code, based on the saturation activity per target nucleus values of the activation detectors irradiated. (author)

  6. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...

  7. Validation of thermalhydraulic codes

    International Nuclear Information System (INIS)

    Wilkie, D.

    1992-01-01

    Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)

  8. High-frequency Total Focusing Method (TFM) imaging in strongly attenuating materials with the decomposition of the time reversal operator associated with orthogonal coded excitations

    Science.gov (United States)

    Villaverde, Eduardo Lopez; Robert, Sébastien; Prada, Claire

    2017-02-01

    In the present work, the Total Focusing Method (TFM) is used to image defects in a High Density Polyethylene (HDPE) pipe. The viscoelastic attenuation of this material corrupts the images with a high electronic noise. In order to improve the image quality, the Decomposition of the Time Reversal Operator (DORT) filtering is combined with spatial Walsh-Hadamard coded transmissions before calculating the images. Experiments on a complex HDPE joint demonstrate that this method improves the signal-to-noise ratio by more than 40 dB in comparison with the conventional TFM.

  9. A rational method to evaluate tornado-borne missile speed in nuclear power plants. Validation of a numerical code based on Fujita's tornado model

    International Nuclear Information System (INIS)

    Eguchi, Yuzuru; Sugimoto, Soichiro; Hattori, Yasuo; Hirakuchi, Hiromaru

    2015-01-01

    Explanation is given about a rational method to evaluate tornado-borne missile speed, flight distance and flight height to be used for safety design of a nuclear power plant. In the method, the authors employed Fujita's DBT-77 model as a tornado wind model to take the near-ground tornado wind profile into account. A liftoff model of an object on the ground was developed by conservatively modeling the lift force due to ground effect. The wind field model and the liftoff model have been compiled together with a conventional flight model into a computer code, named TONBOS. In this study, especially, the code is verified for one- and two-dimensional free-fall problems as well as a case of 1957 Dallas tornado wind field model, whose solutions are theoretically or numerically known. Finally, the code is validated by typical car behaviors characterized by tornado wind speeds of the enhanced Fujita scale, as well as by an actual event where a truck was blown away by a tornado which struck a part of the town of Saroma, Hokkaido in November, 2006. (author)

  10. WIMS-AECL/RFSP code validation of reactivity calculations following a long shutdown using the simple-cell history-based method

    International Nuclear Information System (INIS)

    Ardeshiri, F.; Donnelly, J.V.; Arsenault, B.

    1998-01-01

    The purpose of this analysis is to validate the Reactor Fuelling Simulation Program (RFSP) using the simple-cell model (SCM) history-based method in a startup simulation following a reactor shutdown period. This study is part of the validation work for history-based calculations, using the WIMS-AECL code with the ENDF/B-V library, and the SCM linked to the RFSP code. In this work, the RFSP code with the SCM history-based method was used to track a 1-year period of the Point Lepreau reactor operating history, that included a 12-day reactor shutdown and subsequent startup. Measured boron and gadolinium concentrations were used in the RFSP simulations, and the predicted values of core reactivity were compared to the reference (pre-shutdown) value. The discrepancies in core reactivity are shown to be better than ±2 milli-k at any time, and better than about ±0.5 milli-k towards the end of the startup transient. The results of this analysis also show that the calculated maximum channel and bundle powers are within an acceptable range during both the core-follow and the reactor startup simulations. (author)

  11. Study of coherent Synchrotron Radiation effects by means of a new simulation code based on the non-linear extension of the operator splitting method

    International Nuclear Information System (INIS)

    Dattoli, G.; Schiavi, A.; Migliorati, M.

    2006-03-01

    The coherent synchrotron radiation (CSR) is one of the main problems limiting the performance of high intensity electron accelerators. The complexity of the physical mechanisms underlying the onset of instabilities due to CSR demands for accurate descriptions, capable of including the large number of features of an actual accelerating device. A code devoted to the analysis of this type of problems should be fast and reliable, conditions that are usually hardly achieved at the same rime. In the past, codes based on Lie algebraic techniques , have been very efficient to treat transport problems in accelerators. The extension of these methods to the non-linear case is ideally suited to treat CSR instability problems. We report on the development of a numerical code, based on the solution of the Vlasov equation, with the inclusion of non-linear contribution due to wake field effects. The proposed solution method exploits an algebraic technique, using exponential operators. We show that the integration procedure is capable of reproducing the onset of an instability and the effects associated with bunching mechanisms leading to the growth of the instability itself. In addition, considerations on the threshold of the instability are also developed [it

  12. Study of coherent synchrotron radiation effects by means of a new simulation code based on the non-linear extension of the operator splitting method

    International Nuclear Information System (INIS)

    Dattoli, G.; Migliorati, M.; Schiavi, A.

    2007-01-01

    The coherent synchrotron radiation (CSR) is one of the main problems limiting the performance of high-intensity electron accelerators. The complexity of the physical mechanisms underlying the onset of instabilities due to CSR demands for accurate descriptions, capable of including the large number of features of an actual accelerating device. A code devoted to the analysis of these types of problems should be fast and reliable, conditions that are usually hardly achieved at the same time. In the past, codes based on Lie algebraic techniques have been very efficient to treat transport problems in accelerators. The extension of these methods to the non-linear case is ideally suited to treat CSR instability problems. We report on the development of a numerical code, based on the solution of the Vlasov equation, with the inclusion of non-linear contribution due to wake field effects. The proposed solution method exploits an algebraic technique that uses the exponential operators. We show that the integration procedure is capable of reproducing the onset of instability and the effects associated with bunching mechanisms leading to the growth of the instability itself. In addition, considerations on the threshold of the instability are also developed

  13. ANISN-L, a CDC-7600 code which solves the one-dimensional, multigroup, time dependent transport equation by the method of discrete ordinates

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, T. P.

    1973-09-20

    The code ANISN-L solves the one-dimensional, multigroup, time-independent Boltzmann transport equation by the method of discrete ordinates. In problems involving a fissionable system, it can calculate the system multiplication or alpha. In such cases, it is also capable of determining isotopic concentrations, radii, zone widths, or buckling in order to achieve a given multiplication or alpha. The code may also calculate fluxes caused by a specified fixed source. Neutron, gamma, and coupled neutron--gamma problems may be solved in either the forward or adjoint (backward) modes. Cross sections describing upscatter, as well as the usual downscatter, may be employed. This report describes the use of ANISN-L; this is a revised version of ANISN which handles both large and small problems efficiently on CDC-7600 computers. (RWR)

  14. Bandwidth efficient coding

    CERN Document Server

    Anderson, John B

    2017-01-01

    Bandwidth Efficient Coding addresses the major challenge in communication engineering today: how to communicate more bits of information in the same radio spectrum. Energy and bandwidth are needed to transmit bits, and bandwidth affects capacity the most. Methods have been developed that are ten times as energy efficient at a given bandwidth consumption as simple methods. These employ signals with very complex patterns and are called "coding" solutions. The book begins with classical theory before introducing new techniques that combine older methods of error correction coding and radio transmission in order to create narrowband methods that are as efficient in both spectrum and energy as nature allows. Other topics covered include modulation techniques such as CPM, coded QAM and pulse design.

  15. Method for measuring the focal spot size of an x-ray tube using a coded aperture mask and a digital detector.

    Science.gov (United States)

    Russo, Paolo; Mettivier, Giovanni

    2011-04-01

    The goal of this study is to evaluate a new method based on a coded aperture mask combined with a digital x-ray imaging detector for measurements of the focal spot sizes of diagnostic x-ray tubes. Common techniques for focal spot size measurements employ a pinhole camera, a slit camera, or a star resolution pattern. The coded aperture mask is a radiation collimator consisting of a large number of apertures disposed on a predetermined grid in an array, through which the radiation source is imaged onto a digital x-ray detector. The method of the coded mask camera allows one to obtain a one-shot accurate and direct measurement of the two dimensions of the focal spot (like that for a pinhole camera) but at a low tube loading (like that for a slit camera). A large number of small apertures in the coded mask operate as a "multipinhole" with greater efficiency than a single pinhole, but keeping the resolution of a single pinhole. X-ray images result from the multiplexed output on the detector image plane of such a multiple aperture array, and the image of the source is digitally reconstructed with a deconvolution algorithm. Images of the focal spot of a laboratory x-ray tube (W anode: 35-80 kVp; focal spot size of 0.04 mm) were acquired at different geometrical magnifications with two different types of digital detector (a photon counting hybrid silicon pixel detector with 0.055 mm pitch and a flat panel CMOS digital detector with 0.05 mm pitch) using a high resolution coded mask (type no-two-holes-touching modified uniformly redundant array) with 480 0.07 mm apertures, designed for imaging at energies below 35 keV. Measurements with a slit camera were performed for comparison. A test with a pinhole camera and with the coded mask on a computed radiography mammography unit with 0.3 mm focal spot was also carried out. The full width at half maximum focal spot sizes were obtained from the line profiles of the decoded images, showing a focal spot of 0.120 mm x 0.105 mm at 35

  16. Decoding Xing-Ling codes

    DEFF Research Database (Denmark)

    Nielsen, Rasmus Refslund

    2002-01-01

    This paper describes an efficient decoding method for a recent construction of good linear codes as well as an extension to the construction. Furthermore, asymptotic properties and list decoding of the codes are discussed.......This paper describes an efficient decoding method for a recent construction of good linear codes as well as an extension to the construction. Furthermore, asymptotic properties and list decoding of the codes are discussed....

  17. An efficient and novel computation method for simulating diffraction patterns from large-scale coded apertures on large-scale focal plane arrays

    Science.gov (United States)

    Shrekenhamer, Abraham; Gottesman, Stephen R.

    2012-10-01

    A novel and memory efficient method for computing diffraction patterns produced on large-scale focal planes by largescale Coded Apertures at wavelengths where diffraction effects are significant has been developed and tested. The scheme, readily implementable on portable computers, overcomes the memory limitations of present state-of-the-art simulation codes such as Zemax. The method consists of first calculating a set of reference complex field (amplitude and phase) patterns on the focal plane produced by a single (reference) central hole, extending to twice the focal plane array size, with one such pattern for each Line-of-Sight (LOS) direction and wavelength in the scene, and with the pattern amplitude corresponding to the square-root of the spectral irradiance from each such LOS direction in the scene at selected wavelengths. Next the set of reference patterns is transformed to generate pattern sets for other holes. The transformation consists of a translational pattern shift corresponding to each hole's position offset and an electrical phase shift corresponding to each hole's position offset and incoming radiance's direction and wavelength. The set of complex patterns for each direction and wavelength is then summed coherently and squared for each detector to yield a set of power patterns unique for each direction and wavelength. Finally the set of power patterns is summed to produce the full waveband diffraction pattern from the scene. With this tool researchers can now efficiently simulate diffraction patterns produced from scenes by large-scale Coded Apertures onto large-scale focal plane arrays to support the development and optimization of coded aperture masks and image reconstruction algorithms.

  18. Phonological coding during reading.

    Science.gov (United States)

    Leinenger, Mallorie

    2014-11-01

    The exact role that phonological coding (the recoding of written, orthographic information into a sound based code) plays during silent reading has been extensively studied for more than a century. Despite the large body of research surrounding the topic, varying theories as to the time course and function of this recoding still exist. The present review synthesizes this body of research, addressing the topics of time course and function in tandem. The varying theories surrounding the function of phonological coding (e.g., that phonological codes aid lexical access, that phonological codes aid comprehension and bolster short-term memory, or that phonological codes are largely epiphenomenal in skilled readers) are first outlined, and the time courses that each maps onto (e.g., that phonological codes come online early [prelexical] or that phonological codes come online late [postlexical]) are discussed. Next the research relevant to each of these proposed functions is reviewed, discussing the varying methodologies that have been used to investigate phonological coding (e.g., response time methods, reading while eye-tracking or recording EEG and MEG, concurrent articulation) and highlighting the advantages and limitations of each with respect to the study of phonological coding. In response to the view that phonological coding is largely epiphenomenal in skilled readers, research on the use of phonological codes in prelingually, profoundly deaf readers is reviewed. Finally, implications for current models of word identification (activation-verification model, Van Orden, 1987; dual-route model, e.g., M. Coltheart, Rastle, Perry, Langdon, & Ziegler, 2001; parallel distributed processing model, Seidenberg & McClelland, 1989) are discussed. (PsycINFO Database Record (c) 2014 APA, all rights reserved).

  19. Modeling and commissioning of a Clinac 600 CD by Monte Carlo method using the BEAMnrc and DOSXYZnrc codes

    Energy Technology Data Exchange (ETDEWEB)

    Junior, Reginaldo G., E-mail: reginaldo.junior@ifmg.edu.br [Instituto Federal de Minas Gerais (IFMG), Formiga, MG (Brazil). Departamento de Engenharia Eletrica; Oliveira, Arno H. de; Sousa, Romulo V., E-mail: arnoheeren@gmail.com, E-mail: romuloverdolin@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Mourao, Arnaldo P., E-mail: apratabhz@gmail.com [Centro Federal de Educacao Tecnologica de Minas Gerais, Belo Horizonte, MG (Brazil)

    2015-07-01

    This paper reports the modeling of a linear accelerator Clinac 600 CD with BEAMnrc application, derived from EGSnrc radiation transport code, indicating relevant details of modeling that traditionally involve difficulties imposed on the process. This accelerator was commissioned by the confrontation of experimental dosimetric data with the computer data obtained by DOSXYZnrc application. The information compared in dosimetry process were: field profiles and dose percentage curves obtained in a water phantom with cubic edge of 30 cm. In all comparisons made, the computational data showed satisfactory precision and discrepancies with the experimental data did not exceed 3%, proving the electiveness of the model. Both the accelerator model and the computational dosimetry methodology, revealed the need for adjustments that probably will allow obtaining more accurate data than those obtained in the simulations presented here. These adjustments are mainly associated to improve the resolution of the eld profiles, the voxelization in phantom and optimization of computing time. (author)

  20. Quasi-optical converters for high-power gyrotrons: a brief review of physical models, numerical methods and computer codes

    International Nuclear Information System (INIS)

    Sabchevski, S; Zhelyazkov, I; Benova, E; Atanassov, V; Dankov, P; Thumm, M; Arnold, A; Jin, J; Rzesnicki, T

    2006-01-01

    Quasi-optical (QO) mode converters are used to transform electromagnetic waves of complex structure and polarization generated in gyrotron cavities into a linearly polarized, Gaussian-like beam suitable for transmission. The efficiency of this conversion as well as the maintenance of low level of diffraction losses are crucial for the implementation of powerful gyrotrons as radiation sources for electron-cyclotron-resonance heating of fusion plasmas. The use of adequate physical models, efficient numerical schemes and up-to-date computer codes may provide the high accuracy necessary for the design and analysis of these devices. In this review, we briefly sketch the most commonly used QO converters, the mathematical base they have been treated on and the basic features of the numerical schemes used. Further on, we discuss the applicability of several commercially available and free software packages, their advantages and drawbacks, for solving QO related problems

  1. Looking back on 10 years of the ATLAS Metadata Interface. Reflections on architecture, code design and development methods

    International Nuclear Information System (INIS)

    Fulachier, J; Albrand, S; Lambert, F; Aidel, O

    2014-01-01

    The 'ATLAS Metadata Interface' framework (AMI) has been developed in the context of ATLAS, one of the largest scientific collaborations. AMI can be considered to be a mature application, since its basic architecture has been maintained for over 10 years. In this paper we describe briefly the architecture and the main uses of the framework within the experiment (TagCollector for release management and Dataset Discovery). These two applications, which share almost 2000 registered users, are superficially quite different, however much of the code is shared and they have been developed and maintained over a decade almost completely by the same team of 3 people. We discuss how the architectural principles established at the beginning of the project have allowed us to continue both to integrate the new technologies and to respond to the new metadata use cases which inevitably appear over such a time period.

  2. Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source

    International Nuclear Information System (INIS)

    Spyrou, N.M.; Awotwi-Pratt, J.B.; Williams, A.M.

    2004-01-01

    An activation analysis facility based on an isotopic neutron source (185 GBq 241 Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, it was attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system. (author)

  3. Analysis of latent variance reduction methods in phase space Monte Carlo calculations for 6, 10 and 18 MV photons by using MCNP code

    International Nuclear Information System (INIS)

    Ezzati, A.O.; Sohrabpour, M.

    2013-01-01

    In this study, azimuthal particle redistribution (APR), and azimuthal particle rotational splitting (APRS) methods are implemented in MCNPX2.4 source code. First of all, the efficiency of these methods was compared to two tallying methods. The APRS is more efficient than the APR method in track length estimator tallies. However in the energy deposition tally, both methods have nearly the same efficiency. Latent variance reduction factors were obtained for 6, 10 and 18 MV photons as well. The APRS relative efficiency contours were obtained. These obtained contours reveal that by increasing the photon energies, the contours depth and the surrounding areas were further increased. The relative efficiency contours indicated that the variance reduction factor is position and energy dependent. The out of field voxels relative efficiency contours showed that latent variance reduction methods increased the Monte Carlo (MC) simulation efficiency in the out of field voxels. The APR and APRS average variance reduction factors had differences less than 0.6% for splitting number of 1000. -- Highlights: ► The efficiency of APR and APRS methods was compared to two tallying methods. ► The APRS is more efficient than the APR method in track length estimator tallies. ► In the energy deposition tally, both methods have nearly the same efficiency. ► Variance reduction factors of these methods are position and energy dependent.

  4. Benchmark of the non-parametric Bayesian deconvolution method implemented in the SINBAD code for X/γ rays spectra processing

    Energy Technology Data Exchange (ETDEWEB)

    Rohée, E. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Coulon, R., E-mail: romain.coulon@cea.fr [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Carrel, F. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Dautremer, T.; Barat, E.; Montagu, T. [CEA, LIST, Laboratoire de Modélisation et Simulation des Systèmes, F-91191 Gif-sur-Yvette (France); Normand, S. [CEA, DAM, Le Ponant, DPN/STXN, F-75015 Paris (France); Jammes, C. [CEA, DEN, Cadarache, DER/SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France)

    2016-11-11

    Radionuclide identification and quantification are a serious concern for many applications as for in situ monitoring at nuclear facilities, laboratory analysis, special nuclear materials detection, environmental monitoring, and waste measurements. High resolution gamma-ray spectrometry based on high purity germanium diode detectors is the best solution available for isotopic identification. Over the last decades, methods have been developed to improve gamma spectra analysis. However, some difficulties remain in the analysis when full energy peaks are folded together with high ratio between their amplitudes, and when the Compton background is much larger compared to the signal of a single peak. In this context, this study deals with the comparison between a conventional analysis based on “iterative peak fitting deconvolution” method and a “nonparametric Bayesian deconvolution” approach developed by the CEA LIST and implemented into the SINBAD code. The iterative peak fit deconvolution is used in this study as a reference method largely validated by industrial standards to unfold complex spectra from HPGe detectors. Complex cases of spectra are studied from IAEA benchmark protocol tests and with measured spectra. The SINBAD code shows promising deconvolution capabilities compared to the conventional method without any expert parameter fine tuning.

  5. Entanglement-assisted quantum MDS codes constructed from negacyclic codes

    Science.gov (United States)

    Chen, Jianzhang; Huang, Yuanyuan; Feng, Chunhui; Chen, Riqing

    2017-12-01

    Recently, entanglement-assisted quantum codes have been constructed from cyclic codes by some scholars. However, how to determine the number of shared pairs required to construct entanglement-assisted quantum codes is not an easy work. In this paper, we propose a decomposition of the defining set of negacyclic codes. Based on this method, four families of entanglement-assisted quantum codes constructed in this paper satisfy the entanglement-assisted quantum Singleton bound, where the minimum distance satisfies q+1 ≤ d≤ n+2/2. Furthermore, we construct two families of entanglement-assisted quantum codes with maximal entanglement.

  6. Rate-adaptive BCH codes for distributed source coding

    DEFF Research Database (Denmark)

    Salmistraro, Matteo; Larsen, Knud J.; Forchhammer, Søren

    2013-01-01

    This paper considers Bose-Chaudhuri-Hocquenghem (BCH) codes for distributed source coding. A feedback channel is employed to adapt the rate of the code during the decoding process. The focus is on codes with short block lengths for independently coding a binary source X and decoding it given its...... strategies for improving the reliability of the decoded result are analyzed, and methods for estimating the performance are proposed. In the analysis, noiseless feedback and noiseless communication are assumed. Simulation results show that rate-adaptive BCH codes achieve better performance than low...... correlated side information Y. The proposed codes have been analyzed in a high-correlation scenario, where the marginal probability of each symbol, Xi in X, given Y is highly skewed (unbalanced). Rate-adaptive BCH codes are presented and applied to distributed source coding. Adaptive and fixed checking...

  7. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  8. Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

    Energy Technology Data Exchange (ETDEWEB)

    Williams, P. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dickson, T. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yin, S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2007-12-01

    The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

  9. The fourth research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    The fourth Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effect' was held during 19-23 May, 2003 in Obninsk, Russian Federation. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first RCM took place in Vienna on 24 - 26 November 1999. The meeting was attended by 19 participants from 7 Member States and one from an international organization (France, Germany, India, Japan, Rep. of Korea, Russian Federation, the United Kingdom, and IAEA). The participants from two Member States (China and the U.S.A.) provided their results and presentation materials even though being absent at the meeting. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN- 600 core were evaluated. Contributions of the participants in the benchmark analyses is shown. This report first addresses the benchmark definitions and specifications given for each Phase and briefly introduces the basic data, computer codes, and methodologies applied to the benchmark analyses by various participants. Then, the results obtained by the participants in terms of calculational uncertainty and their effect on the core transient behavior are intercompared. Finally it addresses some conclusions drawn in the benchmarks

  10. The Physician Recommendation Coding System (PhyReCS): A Reliable and Valid Method to Quantify the Strength of Physician Recommendations During Clinical Encounters.

    Science.gov (United States)

    Scherr, Karen A; Fagerlin, Angela; Williamson, Lillie D; Davis, J Kelly; Fridman, Ilona; Atyeo, Natalie; Ubel, Peter A

    2017-01-01

    Physicians' recommendations affect patients' treatment choices. However, most research relies on physicians' or patients' retrospective reports of recommendations, which offer a limited perspective and have limitations such as recall bias. To develop a reliable and valid method to measure the strength of physician recommendations using direct observation of clinical encounters. Clinical encounters (n = 257) were recorded as part of a larger study of prostate cancer decision making. We used an iterative process to create the 5-point Physician Recommendation Coding System (PhyReCS). To determine reliability, research assistants double-coded 50 transcripts. To establish content validity, we used 1-way analyses of variance to determine whether relative treatment recommendation scores differed as a function of which treatment patients received. To establish concurrent validity, we examined whether patients' perceived treatment recommendations matched our coded recommendations. The PhyReCS was highly reliable (Krippendorf's alpha = 0.89, 95% CI [0.86, 0.91]). The average relative treatment recommendation score for each treatment was higher for individuals who received that particular treatment. For example, the average relative surgery recommendation score was higher for individuals who received surgery versus radiation (mean difference = 0.98, SE = 0.18, P recommendations matched coded recommendations 81% of the time. The PhyReCS is a reliable and valid way to capture the strength of physician recommendations. We believe that the PhyReCS would be helpful for other researchers who wish to study physician recommendations, an important part of patient decision making. © The Author(s) 2016.

  11. Extension of BEPU methods to Sub-channel Thermal-Hydraulics and to Coupled Three-Dimensional Neutronics/Thermal-Hydraulics Codes

    International Nuclear Information System (INIS)

    Avramova, M.; Ivanov, K.; Arenas, C.

    2013-01-01

    The principles that support the risk-informed regulation are to be considered in an integrated decision-making process. Thus, any evaluation of licensing issues supported by a safety analysis would take into account both deterministic and probabilistic aspects of the problem. The deterministic aspects will be addressed using Best Estimate code calculations and considering the associated uncertainties i.e. Plus Uncertainty (BEPU) calculations. In recent years there has been an increasing demand from nuclear research, industry, safety and regulation for best estimate predictions to be provided with their confidence bounds. This applies also to the sub-channel thermal-hydraulic codes, which are used to evaluate local safety parameters. The paper discusses the extension of BEPU methods to the sub-channel thermal-hydraulic codes on the example of the Pennsylvania State University (PSU) version of COBRA-TF (CTF). The use of coupled codes supplemented with uncertainty analysis allows to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1000, EPR-1600 and etc.). The paper presents the application of Generalized Perturbation Theory (GPT) to generate uncertainties associated with the few-group assembly homogenized neutron cross-section data used as input in coupled reactor core calculations. This is followed by a discussion of uncertainty propagation methodologies, being implemented by PSU in cooperation of Technical University of Catalonia (UPC) for reactor core calculations and for comprehensive multi-physics simulations. (authors)

  12. Discrete probability models and methods probability on graphs and trees, Markov chains and random fields, entropy and coding

    CERN Document Server

    Brémaud, Pierre

    2017-01-01

    The emphasis in this book is placed on general models (Markov chains, random fields, random graphs), universal methods (the probabilistic method, the coupling method, the Stein-Chen method, martingale methods, the method of types) and versatile tools (Chernoff's bound, Hoeffding's inequality, Holley's inequality) whose domain of application extends far beyond the present text. Although the examples treated in the book relate to the possible applications, in the communication and computing sciences, in operations research and in physics, this book is in the first instance concerned with theory. The level of the book is that of a beginning graduate course. It is self-contained, the prerequisites consisting merely of basic calculus (series) and basic linear algebra (matrices). The reader is not assumed to be trained in probability since the first chapters give in considerable detail the background necessary to understand the rest of the book. .

  13. Development of a first-principles code based on the screened KKR method for large super-cells

    International Nuclear Information System (INIS)

    Doi, S; Ogura, M; Akai, H

    2013-01-01

    The procedures of performing first-principles electronic structure calculation using the Korringa-Kohn-Rostoker (KKR) and the screened KKR methods are reviewed with an emphasis put on their numerical efficiency. It is shown that an iterative matrix inversion combined with a suitable preconditioning greatly improves the computational time of screened KKR method. The method is well parallelized and also has an O(N) scaling property

  14. MO-F-CAMPUS-I-04: Characterization of Fan Beam Coded Aperture Coherent Scatter Spectral Imaging Methods for Differentiation of Normal and Neoplastic Breast Structures

    Energy Technology Data Exchange (ETDEWEB)

    Morris, R; Albanese, K; Lakshmanan, M; Greenberg, J; Kapadia, A [Duke University Medical Center, Durham, NC, Carl E Ravin Advanced Imaging Laboratories, Durham, NC (United States)

    2015-06-15

    Purpose: This study intends to characterize the spectral and spatial resolution limits of various fan beam geometries for differentiation of normal and neoplastic breast structures via coded aperture coherent scatter spectral imaging techniques. In previous studies, pencil beam raster scanning methods using coherent scatter computed tomography and selected volume tomography have yielded excellent results for tumor discrimination. However, these methods don’t readily conform to clinical constraints; primarily prolonged scan times and excessive dose to the patient. Here, we refine a fan beam coded aperture coherent scatter imaging system to characterize the tradeoffs between dose, scan time and image quality for breast tumor discrimination. Methods: An X-ray tube (125kVp, 400mAs) illuminated the sample with collimated fan beams of varying widths (3mm to 25mm). Scatter data was collected via two linear-array energy-sensitive detectors oriented parallel and perpendicular to the beam plane. An iterative reconstruction algorithm yields images of the sample’s spatial distribution and respective spectral data for each location. To model in-vivo tumor analysis, surgically resected breast tumor samples were used in conjunction with lard, which has a form factor comparable to adipose (fat). Results: Quantitative analysis with current setup geometry indicated optimal performance for beams up to 10mm wide, with wider beams producing poorer spatial resolution. Scan time for a fixed volume was reduced by a factor of 6 when scanned with a 10mm fan beam compared to a 1.5mm pencil beam. Conclusion: The study demonstrates the utility of fan beam coherent scatter spectral imaging for differentiation of normal and neoplastic breast tissues has successfully reduced dose and scan times whilst sufficiently preserving spectral and spatial resolution. Future work to alter the coded aperture and detector geometries could potentially allow the use of even wider fans, thereby making coded

  15. Calculating the Mean Amplitude of Glycemic Excursions from Continuous Glucose Data Using an Open-Code Programmable Algorithm Based on the Integer Nonlinear Method.

    Science.gov (United States)

    Yu, Xuefei; Lin, Liangzhuo; Shen, Jie; Chen, Zhi; Jian, Jun; Li, Bin; Xin, Sherman Xuegang

    2018-01-01

    The mean amplitude of glycemic excursions (MAGE) is an essential index for glycemic variability assessment, which is treated as a key reference for blood glucose controlling at clinic. However, the traditional "ruler and pencil" manual method for the calculation of MAGE is time-consuming and prone to error due to the huge data size, making the development of robust computer-aided program an urgent requirement. Although several software products are available instead of manual calculation, poor agreement among them is reported. Therefore, more studies are required in this field. In this paper, we developed a mathematical algorithm based on integer nonlinear programming. Following the proposed mathematical method, an open-code computer program named MAGECAA v1.0 was developed and validated. The results of the statistical analysis indicated that the developed program was robust compared to the manual method. The agreement among the developed program and currently available popular software is satisfied, indicating that the worry about the disagreement among different software products is not necessary. The open-code programmable algorithm is an extra resource for those peers who are interested in the related study on methodology in the future.

  16. A method to optimize the shield compact and lightweight combining the structure with components together by genetic algorithm and MCNP code.

    Science.gov (United States)

    Cai, Yao; Hu, Huasi; Pan, Ziheng; Hu, Guang; Zhang, Tao

    2018-05-17

    To optimize the shield for neutrons and gamma rays compact and lightweight, a method combining the structure and components together was established employing genetic algorithms and MCNP code. As a typical case, the fission energy spectrum of 235 U which mixed neutrons and gamma rays was adopted in this study. Six types of materials were presented and optimized by the method. Spherical geometry was adopted in the optimization after checking the geometry effect. Simulations have made to verify the reliability of the optimization method and the efficiency of the optimized materials. To compare the materials visually and conveniently, the volume and weight needed to build a shield are employed. The results showed that, the composite multilayer material has the best performance. Copyright © 2018 Elsevier Ltd. All rights reserved.

  17. Transport theory and codes

    International Nuclear Information System (INIS)

    Clancy, B.E.

    1986-01-01

    This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on

  18. Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

    International Nuclear Information System (INIS)

    Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

    2006-01-01

    This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

  19. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  20. Aztheca Code

    International Nuclear Information System (INIS)

    Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

    2017-09-01

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  1. Improvement of neutron kinetics module in TRAC-BF1code: one-dimensional nodal collocation method

    Energy Technology Data Exchange (ETDEWEB)

    Jambrina, Ana; Barrachina, Teresa; Miro, Rafael; Verdu, Gumersindo, E-mail: ajambrina@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidade Politecnica de Valencia (UPV), Valencia (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S.L., Madrid (Spain); Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion S.A.U., Madrid (Spain)

    2013-07-01

    The TRAC-BF1 one-dimensional kinetic model is a formulation of the neutron diffusion equation in the two energy groups' approximation, based on the analytical nodal method (ANM). The advantage compared with a zero-dimensional kinetic model is that the axial power profile may vary with time due to thermal-hydraulic parameter changes and/or actions of the control systems but at has the disadvantages that in unusual situations it fails to converge. The nodal collocation method developed for the neutron diffusion equation and applied to the kinetics resolution of TRAC-BF1 thermal-hydraulics, is an adaptation of the traditional collocation methods for the discretization of partial differential equations, based on the development of the solution as a linear combination of analytical functions. It has chosen to use a nodal collocation method based on a development of Legendre polynomials of neutron fluxes in each cell. The qualification is carried out by the analysis of the turbine trip transient from the NEA benchmark in Peach Bottom NPP using both the original 1D kinetics implemented in TRAC-BF1 and the 1D nodal collocation method. (author)

  2. Considering soil-structure interaction effects in the equivalent static analysis method of the Iranian of the Iranian seismic building code

    International Nuclear Information System (INIS)

    Shakib, H.; Dehghani Ashkezari, G.

    2002-01-01

    In this study, based on the equivalent static analysis method of the Iranian seismic code, an algorithm is presented to consider the soil-structure interaction (SSI) effects. Modifications of free field motion and structural properties like period and damping due to soil situation are considered in the proposed algorithm. An increase for fundamental period of structure and a modification (usually increase) for it's effective damping are observed. The increase of period is due to the flexibility of the soil foundation and modification of damping is due to the dissipating energy in soil. In order to propose the relative expressions in the presented algorithm, the soil-structure analyses of 8, 10, 13 and 16 stories frames are carried out. By considering the NEHRP soil-structure interaction algorithm and findings of soil-structure interaction analyses carried out in this study, the algorithm based on the equivalent static analysis method of the Iranian seismic building code to consider the effect of soil-structure interaction

  3. Development of boiling transition analysis code TCAPE-INS/B based on mechanistic methods for BWR fuel bundles. Models and validations with boiling transition experimental data

    International Nuclear Information System (INIS)

    Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

    2003-01-01

    The Boiling Transition (BT) analysis code TCAPE-INS/B based on the mechanistic methods coupled with subchannel analysis has been developed for the evaluation of the integrity of Boiling Water Reactor (BWR) fuel rod bundles under abnormal operations. Objective of the development is the evaluation of the BT without using empirical BT and rewetting correlations needed for different bundle designs in the current analysis methods. TCAPE-INS/B consisted mainly of the drift-flux model, the film flow model, the cross-flow model, the thermal conductivity model and the heat transfer correlations. These models were validated systematically with the experimental data. The accuracy of the prediction for the steady-state Critical Heat Flux (CHF) and the transient temperature of the fuel rod surface after the occurrence of BT were evaluated on the validations. The calculations for the experiments with the single tube and bundles were carried out for the validations of the models incorporated in the code. The results showed that the steady-state CHF was predicted within about 6% average error. In the transient calculations, BT timing and temperature of the fuel rod surface gradient agreed well with experimental results, but rewetting was predicted lately. So, modeling of heat transfer phenomena during post-BT is under modification. (author)

  4. Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

    International Nuclear Information System (INIS)

    Chen Qichang; Wu Hongchun; Cao Liangzhi

    2008-01-01

    A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively

  5. Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

    Energy Technology Data Exchange (ETDEWEB)

    Chen Qichang; Wu Hongchun [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China); Cao Liangzhi [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China)], E-mail: caolz@mail.xjtu.edu.cn

    2008-10-15

    A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively.

  6. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

    a computational and poetic composition for two screens: on one of these, texts and voices are repeated and disrupted by mathematical chaos, together exploring the performativity of code and language; on the other, is a mix of a computer programming syntax and human language. In this sense queer code can...... be understood as both an object and subject of study that intervenes in the world’s ‘becoming' and how material bodies are produced via human and nonhuman practices. Through mixing the natural and computer language, this article presents a script in six parts from a performative lecture for two persons...

  7. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

    NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases

  8. Evaluation of equivalent doses in 18F PET/CT using the Monte Carlo method with MCNPX code

    International Nuclear Information System (INIS)

    Belinato, Walmir; Santos, William Souza; Perini, Ana Paula; Neves, Lucio Pereira; Souza, Divanizia N.

    2017-01-01

    The present work used the Monte Carlo method (MMC), specifically the Monte Carlo NParticle - MCNPX, to simulate the interaction of radiation involving photons and particles, such as positrons and electrons, with virtual adult anthropomorphic simulators on PET / CT scans and to determine absorbed and equivalent doses in adult male and female patients

  9. Quality Improvement of MARS Code and Establishment of Code Coupling

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Kim, Kyung Doo

    2010-04-01

    The improvement of MARS code quality and coupling with regulatory auditing code have been accomplished for the establishment of self-reliable technology based regulatory auditing system. The unified auditing system code was realized also by implementing the CANDU specific models and correlations. As a part of the quality assurance activities, the various QA reports were published through the code assessments. The code manuals were updated and published a new manual which describe the new models and correlations. The code coupling methods were verified though the exercise of plant application. The education-training seminar and technology transfer were performed for the code users. The developed MARS-KS is utilized as reliable auditing tool for the resolving the safety issue and other regulatory calculations. The code can be utilized as a base technology for GEN IV reactor applications

  10. Scoping Evaluation of the IRIS Radiation Environment by Using the FW-CADIS Method and SCALE MAVRIC Code

    International Nuclear Information System (INIS)

    Petrovic, B.

    2008-01-01

    IRIS is an advanced pressurized water reactor of integral configuration. This integral configuration with its relatively large reactor vessel and thick downcomer (1.7 m) results in a significant reduction of radiation field and material activation. It thus enables setting up aggressive dose reduction objectives, but at the same time presents challenges for the shielding analysis which needs to be performed over a large spatial domain and include flux attenuation by many orders of magnitude. The Monte Carlo method enables accurately representing irregular geometry and potential streaming paths, but may require significant computational efforts to reduce statistical uncertainty within the acceptable range. Variance reduction methods do exist, but they are designed to provide results for individual detectors and in limited regions, whereas in the scoping phase of IRIS shielding analysis the results are sought throughout the whole containment. To facilitate such analysis, the SCALE MAVRIC was employed. Based on the recently developed FW-CADIS method, MAVRIC uses forward and adjoint deterministic transport theory calculations to generate effective biasing parameters for Monte Carlo simulations throughout the problem. Previous studies have confirmed the potential of this method for obtaining Monte Carlo solutions with acceptable statistics over large spatial domains. The objective of this work was to evaluate the capability of the FW-CADIS/MAVRIC to efficiently perform the required shielding analysis of IRIS. For that purpose, a representative model was prepared, retaining the main problem characteristics, i.e., a large spatial domain (over 10 m in each dimension) and significant attenuation (over 12 orders of magnitude), but geometrically rather simplified. The obtained preliminary results indicate that the FW-CADIS method implemented through the MAVRIC sequence in SCALE will enable determination of radiation field throughout the large spatial domain of the IRIS nuclear

  11. Benchmark calculations of power distribution within fuel assemblies. Phase 2: comparison of data reduction and power reconstruction methods in production codes

    International Nuclear Information System (INIS)

    2000-01-01

    Systems loaded with plutonium in the form of mixed-oxide (MOX) fuel show somewhat different neutronic characteristics compared with those using conventional uranium fuels. In order to maintain adequate safety standards, it is essential to accurately predict the characteristics of MOX-fuelled systems and to further validate both the nuclear data and the computation methods used. A computation benchmark on power distribution within fuel assemblies to compare different techniques used in production codes for fine flux prediction in systems partially loaded with MOX fuel was carried out at an international level. It addressed first the numerical schemes for pin power reconstruction, then investigated the global performance including cross-section data reduction methods. This report provides the detailed results of this second phase of the benchmark. The analysis of the results revealed that basic data still need to be improved, primarily for higher plutonium isotopes and minor actinides. (author)

  12. Method and codes for solving the optimization problem of initial material distribution and controlling of reactor during the run

    International Nuclear Information System (INIS)

    Isakova, L.Ya.; Rachkova, D.A.; Vtorova, O.Yu.; Matekin, M.P.; Sobol, I.M.

    1992-01-01

    The optimization problem of initial distribution of fuel composition and controlling of the reactor during the run is solved. The optimization problem is formulated as a multicriterial one with different types of constraints. The distinguished feature of the method proposed is the systematic scanning of multidimensional ares, where the trial points in the space of parameters are the points of uniformly distributed LP τ -sequences. The reactor computation is carried out by the four group diffusion method in two-dimensional cylindrical geometry. The burnup absorbers are taken into account as additional absorption cross-sections, represented by approximants. The tables of trials make possible the estimation of the values of global extrema. The coordinates of the points where the external values are attained can be estimated too

  13. Comparison of transform coding methods with an optimal predictor for the data compression of digital elevation models

    Science.gov (United States)

    Lewis, Michael

    1994-01-01

    Statistical encoding techniques enable the reduction of the number of bits required to encode a set of symbols, and are derived from their probabilities. Huffman encoding is an example of statistical encoding that has been used for error-free data compression. The degree of compression given by Huffman encoding in this application can be improved by the use of prediction methods. These replace the set of elevations by a set of corrections that have a more advantageous probability distribution. In particular, the method of Lagrange Multipliers for minimization of the mean square error has been applied to local geometrical predictors. Using this technique, an 8-point predictor achieved about a 7 percent improvement over an existing simple triangular predictor.

  14. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  15. Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the objectives, meeting goals and overall NASA goals for the NASA Data Standards Working Group. The presentation includes information on the technical progress surrounding the objective, short LDPC codes, and the general results on the Pu-Pw tradeoff.

  16. ANIMAL code

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1979-01-01

    This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables

  17. Network Coding

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 15; Issue 7. Network Coding. K V Rashmi Nihar B Shah P Vijay Kumar. General Article Volume 15 Issue 7 July 2010 pp 604-621. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/015/07/0604-0621 ...

  18. Expander Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Expander Codes - The Sipser–Spielman Construction. Priti Shankar. General Article Volume 10 ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science Bangalore 560 012, India.

  19. Report number codes

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.N. (ed.)

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name.

  20. Report number codes

    International Nuclear Information System (INIS)

    Nelson, R.N.

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name