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Sample records for performance test code

  1. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  2. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  3. Test Code Quality and Its Relation to Issue Handling Performance

    NARCIS (Netherlands)

    Athanasiou, D.; Nugroho, A.; Visser, J.; Zaidman, A.

    2014-01-01

    Automated testing is a basic principle of agile development. Its benefits include early defect detection, defect cause localization and removal of fear to apply changes to the code. Therefore, maintaining high quality test code is essential. This study introduces a model that assesses test code

  4. Verification testing of the compression performance of the HEVC screen content coding extensions

    Science.gov (United States)

    Sullivan, Gary J.; Baroncini, Vittorio A.; Yu, Haoping; Joshi, Rajan L.; Liu, Shan; Xiu, Xiaoyu; Xu, Jizheng

    2017-09-01

    This paper reports on verification testing of the coding performance of the screen content coding (SCC) extensions of the High Efficiency Video Coding (HEVC) standard (Rec. ITU-T H.265 | ISO/IEC 23008-2 MPEG-H Part 2). The coding performance of HEVC screen content model (SCM) reference software is compared with that of the HEVC test model (HM) without the SCC extensions, as well as with the Advanced Video Coding (AVC) joint model (JM) reference software, for both lossy and mathematically lossless compression using All-Intra (AI), Random Access (RA), and Lowdelay B (LB) encoding structures and using similar encoding techniques. Video test sequences in 1920×1080 RGB 4:4:4, YCbCr 4:4:4, and YCbCr 4:2:0 colour sampling formats with 8 bits per sample are tested in two categories: "text and graphics with motion" (TGM) and "mixed" content. For lossless coding, the encodings are evaluated in terms of relative bit-rate savings. For lossy compression, subjective testing was conducted at 4 quality levels for each coding case, and the test results are presented through mean opinion score (MOS) curves. The relative coding performance is also evaluated in terms of Bjøntegaard-delta (BD) bit-rate savings for equal PSNR quality. The perceptual tests and objective metric measurements show a very substantial benefit in coding efficiency for the SCC extensions, and provided consistent results with a high degree of confidence. For TGM video, the estimated bit-rate savings ranged from 60-90% relative to the JM and 40-80% relative to the HM, depending on the AI/RA/LB configuration category and colour sampling format.

  5. Refactoring test code

    NARCIS (Netherlands)

    A. van Deursen (Arie); L.M.F. Moonen (Leon); A. van den Bergh; G. Kok

    2001-01-01

    textabstractTwo key aspects of extreme programming (XP) are unit testing and merciless refactoring. Given the fact that the ideal test code / production code ratio approaches 1:1, it is not surprising that unit tests are being refactored. We found that refactoring test code is different from

  6. Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Aakre, Shaun R. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering; Jentz, Ian W. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering; Anderson, Mark H. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering

    2018-03-27

    The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminary observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO2 systems, which commonly employ printed-circuit heat exchangers.

  7. Development of a general coupling interface for the fuel performance code TRANSURANUS – Tested with the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.; Macián-Juan, R.

    2015-01-01

    Highlights: • A general coupling interface was developed for couplings of the TRANSURANUS code. • With this new tool simplified fuel behavior models in codes can be replaced. • Applicable e.g. for several reactor types and from normal operation up to DBA. • The general coupling interface was applied to the reactor dynamics code DYN3D. • The new coupled code system DYN3D–TRANSURANUS was successfully tested for RIA. - Abstract: A general interface is presented for coupling the TRANSURANUS fuel performance code with thermal hydraulics system, sub-channel thermal hydraulics, computational fluid dynamics (CFD) or reactor dynamics codes. As first application the reactor dynamics code DYN3D was coupled at assembly level in order to describe the fuel behavior in more detail. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach transfers parameters like fuel temperature and cladding temperature back to DYN3D. Results of the coupled code system are presented for the reactivity transient scenario, initiated by control rod ejection. More precisely, the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy. These differences can be explained thanks to the greater detail in fuel behavior modeling. The numerical performance for DYN3D–TRANSURANUS was proved to be fast and stable. The coupled code system can therefore improve the assessment of safety criteria, at a reasonable computational cost

  8. Verification of the code ATHLET by post-test analysis of two experiments performed at the CCTF integral test facility

    International Nuclear Information System (INIS)

    Krepper, E.; Schaefer, F.

    2001-03-01

    In the framework of the external validation of the thermohydraulic code ATHLET Mod 1.2 Cycle C, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the japanese test facility CCTF. The test facility CCTF is a 1:25 volume-scaled model of a 1000 MW pressurized water reactor. The tests simulate a double end break in the cold leg of the PWR with ECC injection into the cold leg and with combined ECC injection into the hot and cold legs. The evaluation of the calculated results shows, that the main phenomena can be calculated in a good agreement with the experiment. Especially the behaviour of the quench front and the core cooling are calculated very well. Applying a two-channel representation of the reactor model the radial behaviour of the quench front could be reproduced. Deviations between calculations and experiment can be observed simulating the emergency injection in the beginning of the transient. Very high condensation rates were calculated and the pressure decrease in this phase of the transient is overestimated. Besides that, the pressurization due to evaporation in the refill phase is underestimated by ATHLET. (orig.) [de

  9. Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

    Energy Technology Data Exchange (ETDEWEB)

    Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P. [and others

    1996-12-01

    This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising.

  10. Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

    International Nuclear Information System (INIS)

    Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P.

    1996-01-01

    This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising

  11. Performance Prediction of Centrifugal Compressor for Drop-In Testing Using Low Global Warming Potential Alternative Refrigerants and Performance Test Codes

    Directory of Open Access Journals (Sweden)

    Joo Hoon Park

    2017-12-01

    Full Text Available As environmental regulations to stall global warming are strengthened around the world, studies using newly developed low global warming potential (GWP alternative refrigerants are increasing. In this study, substitute refrigerants, R-1234ze (E and R-1233zd (E, were used in the centrifugal compressor of an R-134a 2-stage centrifugal chiller with a fixed rotational speed. Performance predictions and thermodynamic analyses of the centrifugal compressor for drop-in testing were performed. A performance prediction method based on the existing ASME PTC-10 performance test code was proposed. The proposed method yielded the expected operating area and operating point of the centrifugal compressor with alternative refrigerants. The thermodynamic performance of the first and second stages of the centrifugal compressor was calculated as the polytropic state. To verify the suitability of the proposed method, the drop-in test results of the two alternative refrigerants were compared. The predicted operating range based on the permissible deviation of ASME PTC-10 confirmed that the temperature difference was very small at the same efficiency. Because the drop-in test of R-1234ze (E was performed within the expected operating range, the centrifugal compressor using R-1234ze (E is considered well predicted. However, the predictions of the operating point and operating range of R-1233zd (E were lower than those of the drop-in test. The proposed performance prediction method will assist in understanding thermodynamic performance at the expected operating point and operating area of a centrifugal compressor using alternative gases based on limited design and structure information.

  12. The performance test of anti-scattering x-ray grid with inclined shielding material by MCNP code simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Woo; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-06-15

    The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination.

  13. U3Si2 Fabrication and Testing for Implementation into the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Knight, Travis W.

    2018-04-23

    A creep test stand was designed and constructed for compressive creep testing of U3Si2 pellets. This is described in Chapter 3.

    • Creep testing of U3Si2 pellets was completed. In total, 13 compressive creep tests of U3Si2 pellets was successfully completed. This is reported in Chapter 3.
    • Secondary creep model of U3Si2 was developed and implemented in BISON. This is described in Chapter 4.
    • Properties of U3Si2 were implemented in BISON. This is described in Chapter 4.
    • A resonant frequency and damping analyzer (RFDA) using impulse excitation technique (IET) was setup, tested, and used to analyze U3Si2 samples to measure Young’s and Shear Moduli which were then used to calculate the Poisson ratio for U3Si2. This is described in Chapter 5.
    • Characterization of U3Si2 samples was completed. Samples were prepared and analyzed by XRD, SEM, and optical microscopy. Grain size analysis was conducted on images.
    SEM with EDS was used to analyze second phase precipitates. Impulse excitation technique was used to determine the Young’s and Shear Moduli of a tile specimen which allowed for the determination of the Poisson ratio. Helium pycnometry and mercury intrusion porosimetry was performed and used with image analysis to determine porosity size distribution. Vickers microindentation characterization method was used to evaluate the mechanical properties of U3Si2 including toughness, hardness, and Vickers hardness. Electrical resistivity measurement was done using the four-point probe method. This is reported in Chapter 5.

  14. Development and verification test of integral reactor major components - Development of MCP impeller design, performance prediction code and experimental verification

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Myung Kyoon; Oh, Woo Hyoung; Song, Jae Wook [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    The present study is aimed at developing a computational code for design and performance prediction of an axial-flow pump. The proposed performance prediction method is tested against a model axial-flow pump streamline curvature method. The preliminary design is made by using the ideal velocity triangles at inlet and exit and the three dimensional blade shape is calculated by employing the free vortex design method. Then the detailed blading design is carried out by using experimental database of double circular arc cambered hydrofoils. To computationally determine the design incidence, deviation, blade camber, solidity and stagger angle, a number of correlation equations are developed form the experimental database and a theorical formula for the lift coefficient is adopted. A total of 8 equations are solved iteratively using an under-relaxation factor. An experimental measurement is conducted under a non-cavitating condition to obtain the off-design performance curve and also a cavitation test is carried out by reducing the suction pressure. The experimental results are very satisfactorily compared with the predictions by the streamline curvature method. 28 refs., 26 figs., 11 tabs. (Author)

  15. Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

    2013-01-01

    Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

  16. Smells in software test code

    NARCIS (Netherlands)

    Garousi, Vahid; Küçük, Barış

    2018-01-01

    As a type of anti-pattern, test smells are defined as poorly designed tests and their presence may negatively affect the quality of test suites and production code. Test smells are the subject of active discussions among practitioners and researchers, and various guidelines to handle smells are

  17. Performance Study of Monte Carlo Codes on Xeon Phi Coprocessors — Testing MCNP 6.1 and Profiling ARCHER Geometry Module on the FS7ONNi Problem

    Science.gov (United States)

    Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George

    2017-09-01

    This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

  18. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1994-01-01

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  19. The fuel performance code future

    International Nuclear Information System (INIS)

    Ronchi, C.; Van de Laar, J.

    1988-01-01

    The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

  20. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  1. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  2. An analysis, using the CLAPTRAP code, of the pressure transients developed in the Carolinas Virginia Tube Reactor during containment performance tests

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1982-11-01

    To check containment performance of the CVTR, steam was injected above the operating floor through a 10 foot pipe cap containing the 1 inch diameter holes, at a steady rate of 102.8 lb/sec for a period of 166 seconds. This steam had an enthalpy of 1195 Btu/lb and was therefore not entirely typical of the much wetter material which would be rejected for the greater part of a true breached circuit accident. Pressure transients measured experimentally within the containment were compared with results calculated by the American code CONTEMPT and these results in turn have allowed the Winfrith code CLAPTRAP to be tested for consistency and to establish that the use of this code would have led to similar conclusions about the heat transfer coefficients at the heat absorbent surfaces. (U.K.)

  3. Simulation of buoyancy induced gas mixing tests performed in a large scale containment facility using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Z.; Chin, Y.S. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    This paper compares containment thermal-hydraulics simulations performed using GOTHIC against a past test set of large scale buoyancy induced helium-air-steam mixing experiments that had been performed at the AECL's Chalk River Laboratories. A number of typical post-accident containment phenomena, including thermal/gas stratification, natural convection, cool air entrainment, steam condensation on concrete walls and active local air cooler, were covered. The results provide useful insights into hydrogen gas mixing behaviour following a loss-of-coolant accident and demonstrate GOTHIC's capability in simulating these phenomena. (author)

  4. Simulation of buoyancy induced gas mixing tests performed in a large scale containment facility using GOTHIC code

    International Nuclear Information System (INIS)

    Liang, Z.; Chin, Y.S.

    2014-01-01

    This paper compares containment thermal-hydraulics simulations performed using GOTHIC against a past test set of large scale buoyancy induced helium-air-steam mixing experiments that had been performed at the AECL's Chalk River Laboratories. A number of typical post-accident containment phenomena, including thermal/gas stratification, natural convection, cool air entrainment, steam condensation on concrete walls and active local air cooler, were covered. The results provide useful insights into hydrogen gas mixing behaviour following a loss-of-coolant accident and demonstrate GOTHIC's capability in simulating these phenomena. (author)

  5. Assessment of the prediction capability of the TRANSURANUS fuel performance code on the basis of power ramp tested LWR fuel rods

    International Nuclear Information System (INIS)

    Pastore, G.; Botazzoli, P.; Di Marcello, V.; Luzzi, L.

    2009-01-01

    The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the 'ITU model' for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code. Keywords: Light Water Reactors, Fuel Rod Performance, Power Ramps, Fission Gas Burst Release, Fuel Swelling, Pellet Cladding Interaction, Stress Corrosion Cracking

  6. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  7. Analysis, by RELAP5 code, of boron dilution phenomena in a mid-loop operation transient, performed in PKL III F2.1 RUN 1 test

    International Nuclear Information System (INIS)

    Mascari, F.; Vella, G.; Del Nevo, A.; D'Auria, F.

    2007-01-01

    The present paper deals with the post test analysis and accuracy quantification of the test PKL III F2.1 RUN 1 by RELAP5/Mod3.3 code performed in the framework of the international OECD/SETH PKL III Project. The PKL III is a full-height integral test facility (ITF) that models the entire primary system and most of the secondary system (except for turbine and condenser) of pressurized water reactor of KWU design of the 1300-MW (electric) class on a scale of 1:145. Detailed design was based to the largest possible extent on the specific data of Philippsburg nuclear power plant, unit 2. As for the test facilities of this size, the scaling concept aims to simulate overall thermal hydraulic behavior of the full-scale power plant [1]. The main purpose of the project is to investigate PWR safety issues related to boron dilution and in particular this experiment investigates (a) the boron dilution issue during mid-loop operation and shutdown conditions, and (b) assessing primary circuit accident management operations to prevent boron dilution as a consequence of loss of heat removal [2]. In this work the authors deal with a systematic procedure (developed at the university of Pisa) for code assessment and uncertainty qualification and its application to RELAP5 system code. It is used to evaluate the capability of RELAP5 to reproduce the thermal hydraulics of an inadvertent boron dilution event in a PWR. The quantitative analysis has been performed adopting the Fast Fourier Transform Based Method (FFTBM), which has the capability to quantify the errors in code predictions as compared to the measured experimental signal. (author)

  8. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  9. Assessment of SPACE Code Using the LSTF 10% MSLB Test

    International Nuclear Information System (INIS)

    Kim, Yo Han; Yang, Chang Keun; Ha, Sang Jun

    2012-01-01

    The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a multipurpose nuclear safety analysis code called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors (PWRs). As in the second phase of the project, the beta version of the code has been developed through the validation and verification (V and V) using integral loop test data or plant operating data and the complement of code to solve the SPACE code user problem and resolution reports. In this study, the Large Scale Test Facility (LSTF) 10% main steam line break (MSLB) test, SB-SL-01, was simulated as a V and V work. The results were compared with the experimental data and those of the RELAP5/MOD3.1 code simulation

  10. Standardized Definitions for Code Verification Test Problems

    Energy Technology Data Exchange (ETDEWEB)

    Doebling, Scott William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-14

    This document contains standardized definitions for several commonly used code verification test problems. These definitions are intended to contain sufficient information to set up the test problem in a computational physics code. These definitions are intended to be used in conjunction with exact solutions to these problems generated using Exact- Pack, www.github.com/lanl/exactpack.

  11. Nondestructive testing standards and the ASME code

    International Nuclear Information System (INIS)

    Spanner, J.C.

    1991-04-01

    Nondestructive testing (NDT) requirements and standards are an important part of the ASME Boiler and Pressure Vessel Code. In this paper, the evolution of these requirements and standards is reviewed in the context of the unique technical and legal stature of the ASME Code. The coherent and consistent manner by which the ASME Code rules are organized is described, and the interrelationship between the various ASME Code sections, the piping codes, and the ASTM Standards is discussed. Significant changes occurred in ASME Sections 5 and 11 during the 1980s, and these are highlighted along with projections and comments regarding future trends and changes in these important documents. 4 refs., 8 tabs

  12. SCANAIR: A transient fuel performance code

    International Nuclear Information System (INIS)

    Moal, Alain; Georgenthum, Vincent; Marchand, Olivier

    2014-01-01

    Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

  13. SCANAIR: A transient fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Moal, Alain, E-mail: alain.moal@irsn.fr; Georgenthum, Vincent; Marchand, Olivier

    2014-12-15

    Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

  14. SPACE Code Assessment for FLECHT Test

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Hyoung Kyoun; Min, Ji Hong; Park, Chan Eok; Park, Seok Jeong; Kim, Shin Whan [KEPCO E and C, Daejeon (Korea, Republic of)

    2015-10-15

    According to 10 CFR 50 Appendix K, Emergency Core Cooling System (ECCS) performance evaluation model during LBLOCA should be based on the data of FLECHT test. Heat transfer coefficient (HTC) and Carryout Rate Fraction (CRF) during reflood period of LBLOCA should be conservative. To develop Mass and Energy Release (MER) methodology using Safety and Performance Analysis CodE (SPACE), FLECHT test results were compared to the results calculated by SPACE. FLECHT test facility is modeled to compare the reflood HTC and CRF using SPACE. Sensitivity analysis is performed with various options for HTC correlation. Based on this result, it is concluded that the reflood HTC and CRF calculated with COBRA-TF correlation during LBLOCA meet the requirement of 10 CFR 50 Appendix K. In this study, the analysis results using SPACE predicts heat transfer phenomena of FLECHT test reasonably and conservatively. Reflood HTC for the test number of 0690, 3541 and 4225 are conservative in the reference case. In case of 6948 HTC using COBRATF is conservative to calculate film boiling region. All of analysis results for CRF have sufficient conservatism. Based on these results, it is possible to apply with COBRA-TF correlation to develop MER methodology to analyze LBLOCA using SPACE.

  15. Esophageal function testing: Billing and coding update.

    Science.gov (United States)

    Khan, A; Massey, B; Rao, S; Pandolfino, J

    2018-01-01

    Esophageal function testing is being increasingly utilized in diagnosis and management of esophageal disorders. There have been several recent technological advances in the field to allow practitioners the ability to more accurately assess and treat such conditions, but there has been a relative lack of education in the literature regarding the associated Common Procedural Terminology (CPT) codes and methods of reimbursement. This review, commissioned and supported by the American Neurogastroenterology and Motility Society Council, aims to summarize each of the CPT codes for esophageal function testing and show the trends of associated reimbursement, as well as recommend coding methods in a practical context. We also aim to encourage many of these codes to be reviewed on a gastrointestinal (GI) societal level, by providing evidence of both discrepancies in coding definitions and inadequate reimbursement in this new era of esophageal function testing. © 2017 John Wiley & Sons Ltd.

  16. Modification and testing of the code POLLA

    International Nuclear Information System (INIS)

    Carlson, B.V.; Chalhoub, E.S.; Melnikoff, M.

    1985-01-01

    The implantation and testing of POLLA computer code which translates the paramters of solved resonance for low energy neutrons by Reich-Moore formalism into the equivalent Adler-Adler ones are discussed. The POLLA computer code was developed by Nuclear Data Center of Instituto de Estudos Avancados, in Brazil, to solve actinide resonance cross sections. (Author) [pt

  17. PAPIRUS - a computer code for FBR fuel performance analysis

    International Nuclear Information System (INIS)

    Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

    1991-01-01

    The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

  18. Blood and Books: Performing Code Switching

    Directory of Open Access Journals (Sweden)

    Jeff Friedman

    2008-05-01

    Full Text Available Code switching is a linguistic term that identifies ways individuals use communication modes and registers to negotiate difference in social relations. This essay suggests that arts-based inquiry, in the form of choreography and performance, provides a suitable and efficacious location within which both verbal and nonverbal channels of code switching can be investigated. Blood and Books, a case study of dance choreography within the context of post-colonial Maori performance in Aotearoa/New Zealand, is described and analyzed for its performance of code switching. The essay is framed by a discussion of how arts-based research within tertiary higher education requires careful negotiation in the form of code switching, as performed by the author's reflexive use of vernacular and formal registers in the essay. URN: urn:nbn:de:0114-fqs0802462

  19. Review of SKB's Code Documentation and Testing

    International Nuclear Information System (INIS)

    Hicks, T.W.

    2005-01-01

    SKB is in the process of developing the SR-Can safety assessment for a KBS 3 repository. The assessment will be based on quantitative analyses using a range of computational codes aimed at developing an understanding of how the repository system will evolve. Clear and comprehensive code documentation and testing will engender confidence in the results of the safety assessment calculations. This report presents the results of a review undertaken on behalf of SKI aimed at providing an understanding of how codes used in the SR 97 safety assessment and those planned for use in the SR-Can safety assessment have been documented and tested. Having identified the codes us ed by SKB, several codes were selected for review. Consideration was given to codes used directly in SKB's safety assessment calculations as well as to some of the less visible codes that are important in quantifying the different repository barrier safety functions. SKB's documentation and testing of the following codes were reviewed: COMP23 - a near-field radionuclide transport model developed by SKB for use in safety assessment calculations. FARF31 - a far-field radionuclide transport model developed by SKB for use in safety assessment calculations. PROPER - SKB's harness for executing probabilistic radionuclide transport calculations using COMP23 and FARF31. The integrated analytical radionuclide transport model that SKB has developed to run in parallel with COMP23 and FARF31. CONNECTFLOW - a discrete fracture network model/continuum model developed by Serco Assurance (based on the coupling of NAMMU and NAPSAC), which SKB is using to combine hydrogeological modelling on the site and regional scales in place of the HYDRASTAR code. DarcyTools - a discrete fracture network model coupled to a continuum model, recently developed by SKB for hydrogeological modelling, also in place of HYDRASTAR. ABAQUS - a finite element material model developed by ABAQUS, Inc, which is used by SKB to model repository buffer

  20. Performance measures for transform data coding.

    Science.gov (United States)

    Pearl, J.; Andrews, H. C.; Pratt, W. K.

    1972-01-01

    This paper develops performance criteria for evaluating transform data coding schemes under computational constraints. Computational constraints that conform with the proposed basis-restricted model give rise to suboptimal coding efficiency characterized by a rate-distortion relation R(D) similar in form to the theoretical rate-distortion function. Numerical examples of this performance measure are presented for Fourier, Walsh, Haar, and Karhunen-Loeve transforms.

  1. Transmutation Fuel Performance Code Thermal Model Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller; Pavel G. Medvedev

    2007-09-01

    FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

  2. Performance studies of the parallel VIM code

    International Nuclear Information System (INIS)

    Shi, B.; Blomquist, R.N.

    1996-01-01

    In this paper, the authors evaluate the performance of the parallel version of the VIM Monte Carlo code on the IBM SPx at the High Performance Computing Research Facility at ANL. Three test problems with contrasting computational characteristics were used to assess effects in performance. A statistical method for estimating the inefficiencies due to load imbalance and communication is also introduced. VIM is a large scale continuous energy Monte Carlo radiation transport program and was parallelized using history partitioning, the master/worker approach, and p4 message passing library. Dynamic load balancing is accomplished when the master processor assigns chunks of histories to workers that have completed a previously assigned task, accommodating variations in the lengths of histories, processor speeds, and worker loads. At the end of each batch (generation), the fission sites and tallies are sent from each worker to the master process, contributing to the parallel inefficiency. All communications are between master and workers, and are serial. The SPx is a scalable 128-node parallel supercomputer with high-performance Omega switches of 63 microsec latency and 35 MBytes/sec bandwidth. For uniform and reproducible performance, they used only the 120 identical regular processors (IBM RS/6000) and excluded the remaining eight planet nodes, which may be loaded by other's jobs

  3. Test Driven Development: Performing Art

    Science.gov (United States)

    Bache, Emily

    The art of Test Driven Development (TDD) is a skill that needs to be learnt, and which needs time and practice to master. In this workshop a select number of conference participants with considerable skill and experience are invited to perform code katas [1]. The aim is for them to demonstrate excellence and the use of Test Driven Development, and result in some high quality code. This would be for the benefit of the many programmers attending the conference, who could come along and witness high quality code being written using TDD, and get a chance to ask questions and provide feedback.

  4. Textiles Performance Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

  5. The NMC code: conduct, performance and ethics.

    Science.gov (United States)

    Goldsmith, Jan

    The Code: Standards of Conduct, Performance and Ethics for Nurses and Midwives is a set of key principles that should underpin the practice of all nurses and midwives, and remind them of their professional responsibilities. It is not just a tool used in fitness-to-practise cases--it should be used to guide daily practice for all nurses and midwives. Alongside other standards, guidance and advice from the NMC, the code should be used to support professional development.

  6. Choreographer Pre-Testing Code Analysis and Operational Testing.

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, David J. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Harrison, Christopher B. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Perr, C. W. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Hurd, Steven A [Sandia National Laboratories (SNL-CA), Livermore, CA (United States)

    2014-07-01

    Choreographer is a "moving target defense system", designed to protect against attacks aimed at IP addresses without corresponding domain name system (DNS) lookups. It coordinates actions between a DNS server and a Network Address Translation (NAT) device to regularly change which publicly available IP addresses' traffic will be routed to the protected device versus routed to a honeypot. More details about how Choreographer operates can be found in Section 2: Introducing Choreographer. Operational considerations for the successful deployment of Choreographer can be found in Section 3. The Testing & Evaluation (T&E) for Choreographer involved 3 phases: Pre-testing, Code Analysis, and Operational Testing. Pre-testing, described in Section 4, involved installing and configuring an instance of Choreographer and verifying it would operate as expected for a simple use case. Our findings were that it was simple and straightforward to prepare a system for a Choreographer installation as well as configure Choreographer to work in a representative environment. Code Analysis, described in Section 5, consisted of running a static code analyzer (HP Fortify) and conducting dynamic analysis tests using the Valgrind instrumentation framework. Choreographer performed well, such that only a few errors that might possibly be problematic in a given operating situation were identified. Operational Testing, described in Section 6, involved operating Choreographer in a representative environment created through EmulyticsTM . Depending upon the amount of server resources dedicated to Choreographer vis-á-vis the amount of client traffic handled, Choreographer had varying degrees of operational success. In an environment with a poorly resourced Choreographer server and as few as 50-100 clients, Choreographer failed to properly route traffic over half the time. Yet, with a well-resourced server, Choreographer handled over 1000 clients without missrouting. Choreographer

  7. Cloud Computing for Complex Performance Codes.

    Energy Technology Data Exchange (ETDEWEB)

    Appel, Gordon John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klein, Brandon Thorin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miner, John Gifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-02-01

    This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.

  8. The METEOR/TRANSURANUS fuel performance code

    International Nuclear Information System (INIS)

    Struzik, C.; Guerin, Y.

    1996-01-01

    The first calculations for the FUMEX exercise were performed using version 1.1 of the METEOR/TRANSURANUS code. Since then, important improvements have been implemented on several models. In its present state, the code describes fuel rod behaviour in standard PWR conditions. Its validity extends to UO 2 and MOX fuels clad in Zircaloy-4. Power transient calculations for UO 2 and Gd doped fuel calculations are possible, but further developments are in progress, and the applications will be fully qualified in version 2.0. A considerable effort is made to replace semi-empirical models with models that have a sounder physical basis. (authors). 14 refs

  9. The UK core performance code package

    International Nuclear Information System (INIS)

    Hutt, P.K.; Gaines, N.; McEllin, M.; White, R.J.; Halsall, M.J.

    1991-01-01

    Over the last few years work has been co-ordinated by Nuclear Electric, originally part of the Central Electricity Generating Board, with contributions from the United Kingdom Atomic Energy Authority and British Nuclear Fuels Limited, to produce a generic, easy-to-use and integrated package of core performance codes able to perform a comprehensive range of calculations for fuel cycle design, safety analysis and on-line operational support for Light Water Reactor and Advanced Gas Cooled Reactor plant. The package consists of modern rationalized generic codes for lattice physics (WIMS), whole reactor calculations (PANTHER), thermal hydraulics (VIPRE) and fuel performance (ENIGMA). These codes, written in FORTRAN77, are highly portable and new developments have followed modern quality assurance standards. These codes can all be run ''stand-alone'' but they are also being integrated within a new UNIX-based interactive system called the Reactor Physics Workbench (RPW). The RPW provides an interactive user interface and a sophisticated data management system. It offers quality assurance features to the user and has facilities for defining complex calculational sequences. The Paper reviews the current capabilities of these components, their integration within the package and outlines future developments underway. Finally, the Paper describes the development of an on-line version of this package which is now being commissioned on UK AGR stations. (author)

  10. Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

    International Nuclear Information System (INIS)

    Rizzo, G.; Vella, G.

    2007-01-01

    The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

  11. LFK, FORTRAN Application Performance Test

    International Nuclear Information System (INIS)

    McMahon, F.H.

    1991-01-01

    1 - Description of program or function: LFK, the Livermore FORTRAN Kernels, is a computer performance test that measures a realistic floating-point performance range for FORTRAN applications. Informally known as the Livermore Loops test, the LFK test may be used as a computer performance test, as a test of compiler accuracy (via checksums) and efficiency, or as a hardware endurance test. The LFK test, which focuses on FORTRAN as used in computational physics, measures the joint performance of the computer CPU, the compiler, and the computational structures in units of Mega-flops/sec or Mflops. A C language version of subroutine KERNEL is also included which executes 24 samples of C numerical computation. The 24 kernels are a hydrodynamics code fragment, a fragment from an incomplete Cholesky conjugate gradient code, the standard inner product function of linear algebra, a fragment from a banded linear equations routine, a segment of a tridiagonal elimination routine, an example of a general linear recurrence equation, an equation of state fragment, part of an alternating direction implicit integration code, an integrate predictor code, a difference predictor code, a first sum, a first difference, a fragment from a two-dimensional particle-in-cell code, a part of a one-dimensional particle-in-cell code, an example of how casually FORTRAN can be written, a Monte Carlo search loop, an example of an implicit conditional computation, a fragment of a two-dimensional explicit hydrodynamics code, a general linear recurrence equation, part of a discrete ordinates transport program, a simple matrix calculation, a segment of a Planck distribution procedure, a two-dimensional implicit hydrodynamics fragment, and determination of the location of the first minimum in an array. 2 - Method of solution: CPU performance rates depend strongly on the maturity of FORTRAN compiler machine code optimization. The LFK test-bed executes the set of 24 kernels three times, resetting the DO

  12. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

    LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

  13. Measuring and test equipment control through bar-code technology

    International Nuclear Information System (INIS)

    Crockett, J.D.; Carr, C.C.

    1993-01-01

    Over the past several years, the use, tracking, and documentation of measuring and test equipment (M ampersand TE) has become a major issue. New regulations are forcing companies to develop new policies for providing use history, traceability, and accountability of M ampersand TE. This paper discusses how the Fast Flux Test Facility (FFTF), operated by Westinghouse Hanford Company and located at the Hanford site in Rich- land, Washington, overcame these obstacles by using a computerized system exercising bar-code technology. A data base was developed to identify M ampersand TE containing 33 separate fields, such as manufacturer, model, range, bar-code number, and other pertinent information. A bar-code label was attached to each piece of M ampersand TE. A second data base was created to identify the employee using the M ampersand TE. The fields contained pertinent user information such as name, location, and payroll number. Each employee's payroll number was bar coded and attached to the back of their identification badge. A computer program was developed to automate certain tasks previously performed and tracked by hand. Bar-code technology was combined with this computer program to control the input and distribution of information, eliminate common mistakes, electronically store information, and reduce the time required to check out the M ampersand TE for use

  14. Void fraction prediction of NUPEC PSBT tests by CATHARE code

    International Nuclear Information System (INIS)

    Del Nevo, A.; Michelotti, L.; Moretti, F.; Rozzia, D.; D'Auria, F.

    2011-01-01

    The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, for drawing attention to their weak points, for identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. three-field codes, two-phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The paper reviews the activity carried out by CATHARE2 code on the basis of the subchannel (four test sections) and presents rod bundle (different axial power profile and test sections) experiments available in the database in steady state and transient conditions. The results demonstrate the accuracy of the code in predicting the void fraction in different thermal-hydraulic conditions. The tests are performed varying the pressure, coolant temperature, mass flow and power. Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests. (author)

  15. Pool swell sub-scale testing and code comparison

    International Nuclear Information System (INIS)

    Elisson, K.

    1981-01-01

    The main objective of the experiment was to investigate the pool swell dynamics in general and the forces on the lowered central part of the diaphragm between drywell and wetwell in particular. Apart from the high speed camera pressure transducers and strain gauges were used to monitor the transient. Data was recorded on a 14 channel FM recorder and then digitalised and plotted. In total more than one hundred tests were performed including parametric variations of for example geometry, break flow, initial drywell pressure and initial water level. In parallel to this experiment pool swell calculations have been performed with the computer codes COPTA and STEALTH. COPTA which is a lumped mass code for pressure suppression containment analysis has a slug pool swell mode. STEALTH which is a general purpose lagrangian hydrodynamics code has been used in a 2-D axisymmetric version. The STEALTH code has been used to calculate the radial variations in the vertical displacement and velocity of the pool surface and to predict the load on the lowered central part of the diaphragm. A comparison between the calculations and the experimental data indicates that both codes are sufficiently correct in their description of the pool swell transient. (orig.)

  16. Performance Analysis of CRC Codes for Systematic and Nonsystematic Polar Codes with List Decoding

    Directory of Open Access Journals (Sweden)

    Takumi Murata

    2018-01-01

    Full Text Available Successive cancellation list (SCL decoding of polar codes is an effective approach that can significantly outperform the original successive cancellation (SC decoding, provided that proper cyclic redundancy-check (CRC codes are employed at the stage of candidate selection. Previous studies on CRC-assisted polar codes mostly focus on improvement of the decoding algorithms as well as their implementation, and little attention has been paid to the CRC code structure itself. For the CRC-concatenated polar codes with CRC code as their outer code, the use of longer CRC code leads to reduction of information rate, whereas the use of shorter CRC code may reduce the error detection probability, thus degrading the frame error rate (FER performance. Therefore, CRC codes of proper length should be employed in order to optimize the FER performance for a given signal-to-noise ratio (SNR per information bit. In this paper, we investigate the effect of CRC codes on the FER performance of polar codes with list decoding in terms of the CRC code length as well as its generator polynomials. Both the original nonsystematic and systematic polar codes are considered, and we also demonstrate that different behaviors of CRC codes should be observed depending on whether the inner polar code is systematic or not.

  17. Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1986-01-01

    The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

  18. Test planning and performance

    International Nuclear Information System (INIS)

    Zola, Maurizio

    2001-01-01

    Testing plan should include Safety guide Q4 - Inspection and testing - A testing plan should be prepared including following information: General information (facility name, item or system reference, procurement document reference, document reference number and status, associated procedures and drawings); A sequential listing of all testing activities; Procedure, work instruction, specification or standard to be followed in respect of each operation and test; Acceptance criteria; Identification of who is performing tests; Identification of hold points; Type of records to be prepared for each test; Persons and organizations having authority for final acceptance. Proposed activities sequence is: visual, electrical and mechanical checks; environmental tests (thermal aging, vibrations aging, radioactive aging); performance evaluation in extreme conditions; dynamic tests with functional checks; final electrical and mechanical checks The planning of the tests should always be performed taking into account an interpretative model: a very tight cooperation is advisable between experimental people and numerical people dealing with the analysis of more or less complex models for the seismic assessment of structures and components. Preparatory phase should include the choice of the following items should be agreed upon with the final user of the tests: Excitation points, Excitation types, Excitation amplitude with respect to frequency, Measuring points. Data acquisition, recording and storage, should take into account the characteristics of the successive data processing: to much data can be cumbersome to be processed, but to few data can make unusable the experimental results. The parameters for time history acquisition should be chosen taking into account data processing: for Shock Response Spectrum calculation some special requirements should be met: frequency bounded signal, high frequency sampling, shock noise. For stationary random-like excitation, the sample length

  19. Implementing and Testing the LINTAB, HEATER and PLOTTAB code package

    International Nuclear Information System (INIS)

    Cullen, D.E.; Smith, J.J.

    1987-07-01

    Enclosed is a description of the magnetic tape or floppy diskette containing the LINTAB, HEATER and PLOTTAB code package. In addition detailed information is provided on implementation and testing of these codes. These codes are documented in IAEA-NDS-84. (author)

  20. Performance analysis of WS-EWC coded optical CDMA networks with/without LDPC codes

    Science.gov (United States)

    Huang, Chun-Ming; Huang, Jen-Fa; Yang, Chao-Chin

    2010-10-01

    One extended Welch-Costas (EWC) code family for the wavelength-division-multiplexing/spectral-amplitude coding (WDM/SAC; WS) optical code-division multiple-access (OCDMA) networks is proposed. This system has a superior performance as compared to the previous modified quadratic congruence (MQC) coded OCDMA networks. However, since the performance of such a network is unsatisfactory when the data bit rate is higher, one class of quasi-cyclic low-density parity-check (QC-LDPC) code is adopted to improve that. Simulation results show that the performance of the high-speed WS-EWC coded OCDMA network can be greatly improved by using the LDPC codes.

  1. Testing efficiency transfer codes for equivalence

    International Nuclear Information System (INIS)

    Vidmar, T.; Celik, N.; Cornejo Diaz, N.; Dlabac, A.; Ewa, I.O.B.; Carrazana Gonzalez, J.A.; Hult, M.; Jovanovic, S.; Lepy, M.-C.; Mihaljevic, N.; Sima, O.; Tzika, F.; Jurado Vargas, M.; Vasilopoulou, T.; Vidmar, G.

    2010-01-01

    Four general Monte Carlo codes (GEANT3, PENELOPE, MCNP and EGS4) and five dedicated packages for efficiency determination in gamma-ray spectrometry (ANGLE, DETEFF, GESPECOR, ETNA and EFFTRAN) were checked for equivalence by applying them to the calculation of efficiency transfer (ET) factors for a set of well-defined sample parameters, detector parameters and energies typically encountered in environmental radioactivity measurements. The differences between the results of the different codes never exceeded a few percent and were lower than 2% in the majority of cases.

  2. State of art in FE-based fuel performance codes

    International Nuclear Information System (INIS)

    Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

    2013-01-01

    Fuel performance codes approximate this complex behavior using an axisymmetric, axially-stacked, one-dimensional radial representation to save computation cost. However, the need for improved modeling of PCMI and, particularly, the importance of multidimensional capability for accurate fuel performance simulation has been identified as safety margin decreases. Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed

  3. System Performance and Testing

    NARCIS (Netherlands)

    Frei, U.; Oversloot, H.

    2004-01-01

    This chapter compares and contrasts the system performance of two widely used solar thermal systems using testing and simulation programs. Solar thermal systems are used in many countries for heating domestically used water. In addition to the simple thermosiphon systems, better designed pumped

  4. Validation and testing of the VAM2D computer code

    International Nuclear Information System (INIS)

    Kool, J.B.; Wu, Y.S.

    1991-10-01

    This document describes two modeling studies conducted by HydroGeoLogic, Inc. for the US NRC under contract no. NRC-04089-090, entitled, ''Validation and Testing of the VAM2D Computer Code.'' VAM2D is a two-dimensional, variably saturated flow and transport code, with applications for performance assessment of nuclear waste disposal. The computer code itself is documented in a separate NUREG document (NUREG/CR-5352, 1989). The studies presented in this report involve application of the VAM2D code to two diverse subsurface modeling problems. The first one involves modeling of infiltration and redistribution of water and solutes in an initially dry, heterogeneous field soil. This application involves detailed modeling over a relatively short, 9-month time period. The second problem pertains to the application of VAM2D to the modeling of a waste disposal facility in a fractured clay, over much larger space and time scales and with particular emphasis on the applicability and reliability of using equivalent porous medium approach for simulating flow and transport in fractured geologic media. Reflecting the separate and distinct nature of the two problems studied, this report is organized in two separate parts. 61 refs., 31 figs., 9 tabs

  5. SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

    2012-10-15

    APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

  6. Steady-State Calculation of the ATLAS Test Facility Using the SPACE Code

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Choi, Ki Yong; Kim, Kyung Doo

    2011-01-01

    The Korean nuclear industry is developing a thermalhydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). Several research and industrial organizations including KAERI (Korea Atomic Energy Research Institute) are participating in the collaboration for the development of the SPACE code. One of the main tasks of KAERI is to carry out separate effect tests (SET) and integral effect tests (IET) for code verification and validation (V and V). The IET has been performed with ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation) based on the design features of the APR1400 (Advanced Power Reactor of 1400MWe). In the present work the SPACE code input-deck for ATLAS is developed and used for simulation of the steady-state conditions of ATLAS as a preliminary work for IET V and V of the SPACE code

  7. A fuel performance code TRUST VIc and its validation

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, M; Kogai, T [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

    1997-08-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs.

  8. Performance evaluation based on data from code reviews

    OpenAIRE

    Andrej, Sekáč

    2016-01-01

    Context. Modern code review tools such as Gerrit have made available great amounts of code review data from different open source projects as well as other commercial projects. Code reviews are used to keep the quality of produced source code under control but the stored data could also be used for evaluation of the software development process. Objectives. This thesis uses machine learning methods for an approximation of review expert’s performance evaluation function. Due to limitations in ...

  9. A fuel performance code TRUST VIc and its validation

    International Nuclear Information System (INIS)

    Ishida, M.; Kogai, T.

    1997-01-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs

  10. Input data required for specific performance assessment codes

    International Nuclear Information System (INIS)

    Seitz, R.R.; Garcia, R.S.; Starmer, R.J.; Dicke, C.A.; Leonard, P.R.; Maheras, S.J.; Rood, A.S.; Smith, R.W.

    1992-02-01

    The Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory generated this report on input data requirements for computer codes to assist States and compacts in their performance assessments. This report gives generators, developers, operators, and users some guidelines on what input data is required to satisfy 22 common performance assessment codes. Each of the codes is summarized and a matrix table is provided to allow comparison of the various input required by the codes. This report does not determine or recommend which codes are preferable

  11. The performance testing

    International Nuclear Information System (INIS)

    Mayr, A.

    1975-01-01

    Concerning the time-schedule of reactor performance tests they normally begin when suppliers or constructors have finished construction and made all necessary construction and coordinated tests. If the last-mentioned tests are conducted profoundly, they contribute substantially to a quick and simple carrying-out of the last performance tests and to the general quality of components and systems. At this stage all components of a system should be properly fixed, machinery, instruments and electrical components adjusted and calibrated, all set-points tested, electrical and other supply units in operation or ready to operate and all functions pretested. Just at this stage of the work most of the existing defects and failures of systems can be found. Remembering the fact that the difficulty of operation of complex systems results from detail problems, it is extremely useful to remove all things of this kind as soon as possible, at the latest at this time where it is done easily and normally quickly without influencing start-up-procedures of other systems or even of the total plant. (orig./TK) [de

  12. Validation of a new 39 neutron group self-shielded library based on the nucleonics analysis of the Lotus fusion-fission hybrid test facility performed with the Monte Carlo code

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1985-02-01

    The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)

  13. Predictive Bias and Sensitivity in NRC Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J.; Luscher, Walter G.; Senor, David J.; Cunningham, Mitchel E.; Lanning, Donald D.; Adkins, Harold E.

    2009-10-01

    The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

  14. Performance of code 'FAIR' in IAEA CRP on FUMEX

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Kakodkar, A.

    1996-01-01

    A modern fuel performance analysis code FAIR has been developed for analysing high burnup fuel pins of water/heavy water cooled reactors. The code employs finite element method for modelling thermo mechanical behaviour of fuel pins and mechanistic models for modelling various physical and chemical phenomena affecting the behaviour of nuclear reactor fuel pins. High burnup affects such as pellet thermal conductivity degradation, enhanced fission gas release and radial flux redistribution are incorporated in the code FAIR. The code FAIR is capable of performing statistical analysis of fuel pins using Monte Carlo technique. The code is implemented on BARC parallel processing system ANUPAM. The code has recently participated in an International Atomic Energy Agency (IAEA) coordinated research program (CRP) on fuel modelling at extended burnups (FUMEX). Nineteen agencies from different countries participated in this exercise. In this CRP, spread over a period of three years, a number of high burnup fuel pins irradiated at Halden reactor are analysed. The first phase of the CRP is a blind code comparison exercise, where the computed results are compared with experimental results. The second phase consists of modifications to the code based on the experimental results of first phase and statistical analysis of fuel pins. The performance of the code FAIR in this CRP has been very good. The present report highlights the main features of code FAIR and its performance in the IAEA CRP on FUMEX. 14 refs., 5 tabs., ills

  15. Performance Evaluation of Spectral Amplitude Codes for OCDMA PON

    DEFF Research Database (Denmark)

    Binti Othman, Maisara; Jensen, Jesper Bevensee; Zhang, Xu

    2011-01-01

    the MAI effects in OCDMA. The performance has been characterized through received optical power (ROP) sensitivity and dispersion tolerance assessments. The numerical results show that the ZCC code has a slightly better performance compared to the other two codes for the ROP and similar behavior against...

  16. Understanding protocol performance: impact of test performance.

    Science.gov (United States)

    Turner, Robert G

    2013-01-01

    This is the second of two articles that examine the factors that determine protocol performance. The objective of these articles is to provide a general understanding of protocol performance that can be used to estimate performance, establish limits on performance, decide if a protocol is justified, and ultimately select a protocol. The first article was concerned with protocol criterion and test correlation. It demonstrated the advantages and disadvantages of different criterion when all tests had the same performance. It also examined the impact of increasing test correlation on protocol performance and the characteristics of the different criteria. To examine the impact on protocol performance when individual tests in a protocol have different performance. This is evaluated for different criteria and test correlations. The results of the two articles are combined and summarized. A mathematical model is used to calculate protocol performance for different protocol criteria and test correlations when there are small to large variations in the performance of individual tests in the protocol. The performance of the individual tests that make up a protocol has a significant impact on the performance of the protocol. As expected, the better the performance of the individual tests, the better the performance of the protocol. Many of the characteristics of the different criteria are relatively independent of the variation in the performance of the individual tests. However, increasing test variation degrades some criteria advantages and causes a new disadvantage to appear. This negative impact increases as test variation increases and as more tests are added to the protocol. Best protocol performance is obtained when individual tests are uncorrelated and have the same performance. In general, the greater the variation in the performance of tests in the protocol, the more detrimental this variation is to protocol performance. Since this negative impact is increased as

  17. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

    2012-01-01

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

  18. Analysis of Isp-42, panda test with the spectra code

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.

    2001-01-01

    International Standard Problems (ISP) are organized in order to assess the ability of computer codes to predict the outcome of accidents in Nuclear Power Plants. The ISP-42 test was performed at Paul Scherrer Institute in 1998, as a sequence of six phases, Phase A through F Blind and open calculations of ISP-42 were performed with the computer code SPECTRA for each of the six phases. SPECTRA is a general tool for thermal-hydraulic analyses. Results of blind calculations are in good agreement with experiment. For open calculations several modifications were made in the model. These were mainly corrections of some input errors made in the model used for blind analysis. Some small improvements to the nodalization were made. Results of open calculations are generally closer to the experiment than the blind results. For phase D the containment pressure prediction was somewhat worse in the open calculation. Based on comparisons of blind and open results with experiment several conclusions may be drawn: - use of long ID structures, in contact with pool and atmosphere should be avoided, - PCC units are better represented with larger amount of Control Volumes, - two parallel junctions should be used to represent large openings between vessels, like drywell air line, etc., - careful verification of input decks is needed, - stratification models in SPECTRA are useful for cases with light gas injection; for complex cases a complementary SPECTRA-CFD analysis may be performed. (author)

  19. Verification of the CONPAS (CONtainment Performance Analysis System) code package

    International Nuclear Information System (INIS)

    Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

    1997-09-01

    CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

  20. Evaporation over sump surface in containment studies: code validation on TOSQAN tests

    International Nuclear Information System (INIS)

    Malet, J.; Gelain, T.; Degrees du Lou, O.; Daru, V.

    2011-01-01

    During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on the TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The tests are air-steam tests, as well as tests with other non-condensable gases (He, CO 2 and SF 6 ) under steady and transient conditions. The results show a good agreement between codes and experiments, indicating a good behaviour of the sump models in both codes. (author)

  1. Code cases for implementing risk-based inservice testing in the ASME OM code

    Energy Technology Data Exchange (ETDEWEB)

    Rowley, C.W.

    1996-12-01

    Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices.

  2. Code cases for implementing risk-based inservice testing in the ASME OM code

    International Nuclear Information System (INIS)

    Rowley, C.W.

    1996-01-01

    Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices

  3. State of art in FE-based fuel performance codes

    International Nuclear Information System (INIS)

    Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

    2013-01-01

    Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled, a well known example being the thermomechanical behavior. Adding to this complexity, important aspects of fuel behavior are inherently

  4. Performance of JPEG Image Transmission Using Proposed Asymmetric Turbo Code

    Directory of Open Access Journals (Sweden)

    Siddiqi Mohammad Umar

    2007-01-01

    Full Text Available This paper gives the results of a simulation study on the performance of JPEG image transmission over AWGN and Rayleigh fading channels using typical and proposed asymmetric turbo codes for error control coding. The baseline JPEG algorithm is used to compress a QCIF ( "Suzie" image. The recursive systematic convolutional (RSC encoder with generator polynomials , that is, (13/11 in decimal, and 3G interleaver are used for the typical WCDMA and CDMA2000 turbo codes. The proposed asymmetric turbo code uses generator polynomials , that is, (13/11; 13/9 in decimal, and a code-matched interleaver. The effect of interleaver in the proposed asymmetric turbo code is studied using weight distribution and simulation. The simulation results and performance bound for proposed asymmetric turbo code for the frame length , code rate with Log-MAP decoder over AWGN channel are compared with the typical system. From the simulation results, it is observed that the image transmission using proposed asymmetric turbo code performs better than that with the typical system.

  5. Data processing codes for fatigue and tensile tests

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, Gustavo; Iorio, A.F.; Crespi, J.C.

    1981-01-01

    The processing of fatigue and tensile tests data in order to obtain several parameters of engineering interest requires a considerable effort of numerical calculus. In order to reduce the time spent in this work and to establish standard data processing from a set of similar type tests, it is very advantageous to have a calculation code for running in a computer. Two codes have been developed in FORTRAN language; one of them predicts cyclic properties of materials from the monotonic and incremental or multiple cyclic step tests (ENSPRED CODE), and the other one reduces data coming from strain controlled low cycle fatigue tests (ENSDET CODE). Two examples are included using Zircaloy-4 material from different manufacturers. (author) [es

  6. SITA version 0. A simulation and code testing assistant for TOUGH2 and MARNIE

    Energy Technology Data Exchange (ETDEWEB)

    Seher, Holger; Navarro, Martin

    2016-06-15

    High quality standards have to be met by those numerical codes that are applied in long-term safety assessments for deep geological repositories for radioactive waste. The software environment SITA (''a simulation and code testing assistant for TOUGH2 and MARNIE'') has been developed by GRS in order to perform automated regression testing for the flow and transport simulators TOUGH2 and MARNIE. GRS uses the codes TOUGH2 and MARNIE in order to assess the performance of deep geological repositories for radioactive waste. With SITA, simulation results of TOUGH2 and MARNIE can be compared to analytical solutions and simulations results of other code versions. SITA uses data interfaces to operate with codes whose input and output depends on the code version. The present report is part of a wider GRS programme to assure and improve the quality of TOUGH2 and MARNIE. It addresses users as well as administrators of SITA.

  7. On the performance of diagonal lattice space-time codes

    KAUST Repository

    Abediseid, Walid; Alouini, Mohamed-Slim

    2013-01-01

    There has been tremendous work done on designing space-time codes for the quasi-static multiple-input multiple output (MIMO) channel. All the coding design up-to-date focuses on either high-performance, high rates, low complexity encoding

  8. Performance of Product Codes and Related Structures with Iterated Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn

    2011-01-01

    Several modifications of product codes have been suggested as standards for optical networks. We show that the performance exhibits a threshold that can be estimated from a result about random graphs. For moderate input bit error probabilities, the output error rates for codes of finite length can...

  9. Performance Analysis of Optical Code Division Multiplex System

    Science.gov (United States)

    Kaur, Sandeep; Bhatia, Kamaljit Singh

    2013-12-01

    This paper presents the Pseudo-Orthogonal Code generator for Optical Code Division Multiple Access (OCDMA) system which helps to reduce the need of bandwidth expansion and improve spectral efficiency. In this paper we investigate the performance of multi-user OCDMA system to achieve data rate more than 1 Tbit/s.

  10. On the Performance of the Cache Coding Protocol

    DEFF Research Database (Denmark)

    Maboudi, Behnaz; Sehat, Hadi; Pahlevani, Peyman

    2018-01-01

    Network coding approaches typically consider an unrestricted recoding of coded packets in the relay nodes to increase performance. However, this can expose the system to pollution attacks that cannot be detected during transmission, until the receivers attempt to recover the data. To prevent thes...

  11. Sample test cases using the environmental computer code NECTAR

    International Nuclear Information System (INIS)

    Ponting, A.C.

    1984-06-01

    This note demonstrates a few of the many different ways in which the environmental computer code NECTAR may be used. Four sample test cases are presented and described to show how NECTAR input data are structured. Edited output is also presented to illustrate the format of the results. Two test cases demonstrate how NECTAR may be used to study radio-isotopes not explicitly included in the code. (U.K.)

  12. On the Performance of the Cache Coding Protocol

    Directory of Open Access Journals (Sweden)

    Behnaz Maboudi

    2018-03-01

    Full Text Available Network coding approaches typically consider an unrestricted recoding of coded packets in the relay nodes to increase performance. However, this can expose the system to pollution attacks that cannot be detected during transmission, until the receivers attempt to recover the data. To prevent these attacks while allowing for the benefits of coding in mesh networks, the cache coding protocol was proposed. This protocol only allows recoding at the relays when the relay has received enough coded packets to decode an entire generation of packets. At that point, the relay node recodes and signs the recoded packets with its own private key, allowing the system to detect and minimize the effect of pollution attacks and making the relays accountable for changes on the data. This paper analyzes the delay performance of cache coding to understand the security-performance trade-off of this scheme. We introduce an analytical model for the case of two relays in an erasure channel relying on an absorbing Markov chain and an approximate model to estimate the performance in terms of the number of transmissions before successfully decoding at the receiver. We confirm our analysis using simulation results. We show that cache coding can overcome the security issues of unrestricted recoding with only a moderate decrease in system performance.

  13. User manual for the probabilistic fuel performance code FRP

    International Nuclear Information System (INIS)

    Friis Jensen, J.; Misfeldt, I.

    1980-10-01

    This report describes the use of the probabilistic fuel performance code FRP. Detailed description of both input to and output from the program are given. The use of the program is illustrated by an example. (author)

  14. On the performance of diagonal lattice space-time codes

    KAUST Repository

    Abediseid, Walid

    2013-11-01

    There has been tremendous work done on designing space-time codes for the quasi-static multiple-input multiple output (MIMO) channel. All the coding design up-to-date focuses on either high-performance, high rates, low complexity encoding and decoding, or targeting a combination of these criteria [1]-[9]. In this paper, we analyze in details the performance limits of diagonal lattice space-time codes under lattice decoding. We present both lower and upper bounds on the average decoding error probability. We first derive a new closed-form expression for the lower bound using the so-called sphere lower bound. This bound presents the ultimate performance limit a diagonal lattice space-time code can achieve at any signal-to-noise ratio (SNR). The upper bound is then derived using the union-bound which demonstrates how the average error probability can be minimized by maximizing the minimum product distance of the code. Combining both the lower and the upper bounds on the average error probability yields a simple upper bound on the the minimum product distance that any (complex) lattice code can achieve. At high-SNR regime, we discuss the outage performance of such codes and provide the achievable diversity-multiplexing tradeoff under lattice decoding. © 2013 IEEE.

  15. Performance analysis of LDPC codes on OOK terahertz wireless channels

    International Nuclear Information System (INIS)

    Liu Chun; Wang Chang; Cao Jun-Cheng

    2016-01-01

    Atmospheric absorption, scattering, and scintillation are the major causes to deteriorate the transmission quality of terahertz (THz) wireless communications. An error control coding scheme based on low density parity check (LDPC) codes with soft decision decoding algorithm is proposed to improve the bit-error-rate (BER) performance of an on-off keying (OOK) modulated THz signal through atmospheric channel. The THz wave propagation characteristics and channel model in atmosphere is set up. Numerical simulations validate the great performance of LDPC codes against the atmospheric fading and demonstrate the huge potential in future ultra-high speed beyond Gbps THz communications. (paper)

  16. Performance optimization of spectral amplitude coding OCDMA system using new enhanced multi diagonal code

    Science.gov (United States)

    Imtiaz, Waqas A.; Ilyas, M.; Khan, Yousaf

    2016-11-01

    This paper propose a new code to optimize the performance of spectral amplitude coding-optical code division multiple access (SAC-OCDMA) system. The unique two-matrix structure of the proposed enhanced multi diagonal (EMD) code and effective correlation properties, between intended and interfering subscribers, significantly elevates the performance of SAC-OCDMA system by negating multiple access interference (MAI) and associated phase induce intensity noise (PIIN). Performance of SAC-OCDMA system based on the proposed code is thoroughly analyzed for two detection techniques through analytic and simulation analysis by referring to bit error rate (BER), signal to noise ratio (SNR) and eye patterns at the receiving end. It is shown that EMD code while using SDD technique provides high transmission capacity, reduces the receiver complexity, and provides better performance as compared to complementary subtraction detection (CSD) technique. Furthermore, analysis shows that, for a minimum acceptable BER of 10-9 , the proposed system supports 64 subscribers at data rates of up to 2 Gbps for both up-down link transmission.

  17. Preserving Envelope Efficiency in Performance Based Code Compliance

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, Brian A. [Thornton Energy Consulting (United States); Sullivan, Greg P. [Efficiency Solutions (United States); Rosenberg, Michael I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Baechler, Michael C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-06-20

    The City of Seattle 2012 Energy Code (Seattle 2014), one of the most progressive in the country, is under revision for its 2015 edition. Additionally, city personnel participate in the development of the next generation of the Washington State Energy Code and the International Energy Code. Seattle has pledged carbon neutrality by 2050 including buildings, transportation and other sectors. The United States Department of Energy (DOE), through Pacific Northwest National Laboratory (PNNL) provided technical assistance to Seattle in order to understand the implications of one potential direction for its code development, limiting trade-offs of long-lived building envelope components less stringent than the prescriptive code envelope requirements by using better-than-code but shorter-lived lighting and heating, ventilation, and air-conditioning (HVAC) components through the total building performance modeled energy compliance path. Weaker building envelopes can permanently limit building energy performance even as lighting and HVAC components are upgraded over time, because retrofitting the envelope is less likely and more expensive. Weaker building envelopes may also increase the required size, cost and complexity of HVAC systems and may adversely affect occupant comfort. This report presents the results of this technical assistance. The use of modeled energy code compliance to trade-off envelope components with shorter-lived building components is not unique to Seattle and the lessons and possible solutions described in this report have implications for other jurisdictions and energy codes.

  18. Automated Testing Infrastructure and Result Comparison for Geodynamics Codes

    Science.gov (United States)

    Heien, E. M.; Kellogg, L. H.

    2013-12-01

    The geodynamics community uses a wide variety of codes on a wide variety of both software and hardware platforms to simulate geophysical phenomenon. These codes are generally variants of finite difference or finite element calculations involving Stokes flow or wave propagation. A significant problem is that codes of even low complexity will return different results depending on the platform due to slight differences in hardware, software, compiler, and libraries. Furthermore, changes to the codes during development may affect solutions in unexpected ways such that previously validated results are altered. The Computational Infrastructure for Geodynamics (CIG) is funded by the NSF to enhance the capabilities of the geodynamics community through software development. CIG has recently done extensive work in setting up an automated testing and result validation system based on the BaTLab system developed at the University of Wisconsin, Madison. This system uses 16 variants of Linux and Mac platforms on both 32 and 64-bit processors to test several CIG codes, and has also recently been extended to support testing on the XSEDE TACC (Texas Advanced Computing Center) Stampede cluster. In this work we overview the system design and demonstrate how automated testing and validation occurs and results are reported. We also examine several results from the system from different codes and discuss how changes in compilers and libraries affect the results. Finally we detail some result comparison tools for different types of output (scalar fields, velocity fields, seismogram data), and discuss within what margins different results can be considered equivalent.

  19. Independent verification and validation testing of the FLASH computer code, Versiion 3.0

    International Nuclear Information System (INIS)

    Martian, P.; Chung, J.N.

    1992-06-01

    Independent testing of the FLASH computer code, Version 3.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at various Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Verification tests, and validation tests, were used to determine the operational status of the FLASH computer code. These tests were specifically designed to test: correctness of the FORTRAN coding, computational accuracy, and suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: blind testing, independent applications, and graduated difficulty of test cases. Both quantitative and qualitative testing was performed through evaluating relative root mean square values and graphical comparisons of the numerical, analytical, and experimental data. Four verification test were used to check the computational accuracy and correctness of the FORTRAN coding, and three validation tests were used to check the suitability to simulating actual conditions. These tests cases ranged in complexity from simple 1-D saturated flow to 2-D variably saturated problems. The verification tests showed excellent quantitative agreement between the FLASH results and analytical solutions. The validation tests showed good qualitative agreement with the experimental data. Based on the results of this testing, it was concluded that the FLASH code is a versatile and powerful two-dimensional analysis tool for fluid flow. In conclusion, all aspects of the code that were tested, except for the unit gradient bottom boundary condition, were found to be fully operational and ready for use in hydrological and environmental studies

  20. Assessment of the SPACE Code Using the ATLAS SLB-GB-01 Test

    International Nuclear Information System (INIS)

    Kim, Yo Han; Yang, Chang Keun; Kim, Seyun

    2013-01-01

    The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a safety analysis code, called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) by collaborative works with other Korean nuclear industries. The SPACE is a general-purpose best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient functions and capabilities to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the SPACE code development project, the 2.14 version of the code was released through the successive various V and V works using integral loop test data or plant operating data. In this study, the ATLAS main steam-line break (MSLB) test, SLB-GB-01, was simulated as a V and V work. The results were compared with the measured data. The ATALS MSLB test, SLB-GB-01, was simulated using the SPACE code. The results were compared with experimental data. Through the simulation, it was concluded that the SPACE code can effectively simulate MSLB accidents

  1. FARO base case post-test analysis by COMETA code

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

    1995-09-01

    The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

  2. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  3. Oxide fuel pin transient performance analysis and design with the TEMECH code

    International Nuclear Information System (INIS)

    Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.

    1986-01-01

    The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments

  4. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

  5. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

    International Nuclear Information System (INIS)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-01-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  6. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  7. Performance enhancement of optical code-division multiple-access systems using transposed modified Walsh code

    Science.gov (United States)

    Sikder, Somali; Ghosh, Shila

    2018-02-01

    This paper presents the construction of unipolar transposed modified Walsh code (TMWC) and analysis of its performance in optical code-division multiple-access (OCDMA) systems. Specifically, the signal-to-noise ratio, bit error rate (BER), cardinality, and spectral efficiency were investigated. The theoretical analysis demonstrated that the wavelength-hopping time-spreading system using TMWC was robust against multiple-access interference and more spectrally efficient than systems using other existing OCDMA codes. In particular, the spectral efficiency was calculated to be 1.0370 when TMWC of weight 3 was employed. The BER and eye pattern for the designed TMWC were also successfully obtained using OptiSystem simulation software. The results indicate that the proposed code design is promising for enhancing network capacity.

  8. WOMBAT: A Scalable and High-performance Astrophysical Magnetohydrodynamics Code

    Energy Technology Data Exchange (ETDEWEB)

    Mendygral, P. J.; Radcliffe, N.; Kandalla, K. [Cray Inc., St. Paul, MN 55101 (United States); Porter, D. [Minnesota Supercomputing Institute for Advanced Computational Research, Minneapolis, MN USA (United States); O’Neill, B. J.; Nolting, C.; Donnert, J. M. F.; Jones, T. W. [School of Physics and Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Edmon, P., E-mail: pjm@cray.com, E-mail: nradclif@cray.com, E-mail: kkandalla@cray.com, E-mail: oneill@astro.umn.edu, E-mail: nolt0040@umn.edu, E-mail: donnert@ira.inaf.it, E-mail: twj@umn.edu, E-mail: dhp@umn.edu, E-mail: pedmon@cfa.harvard.edu [Institute for Theory and Computation, Center for Astrophysics, Harvard University, Cambridge, MA 02138 (United States)

    2017-02-01

    We present a new code for astrophysical magnetohydrodynamics specifically designed and optimized for high performance and scaling on modern and future supercomputers. We describe a novel hybrid OpenMP/MPI programming model that emerged from a collaboration between Cray, Inc. and the University of Minnesota. This design utilizes MPI-RMA optimized for thread scaling, which allows the code to run extremely efficiently at very high thread counts ideal for the latest generation of multi-core and many-core architectures. Such performance characteristics are needed in the era of “exascale” computing. We describe and demonstrate our high-performance design in detail with the intent that it may be used as a model for other, future astrophysical codes intended for applications demanding exceptional performance.

  9. WOMBAT: A Scalable and High-performance Astrophysical Magnetohydrodynamics Code

    International Nuclear Information System (INIS)

    Mendygral, P. J.; Radcliffe, N.; Kandalla, K.; Porter, D.; O’Neill, B. J.; Nolting, C.; Donnert, J. M. F.; Jones, T. W.; Edmon, P.

    2017-01-01

    We present a new code for astrophysical magnetohydrodynamics specifically designed and optimized for high performance and scaling on modern and future supercomputers. We describe a novel hybrid OpenMP/MPI programming model that emerged from a collaboration between Cray, Inc. and the University of Minnesota. This design utilizes MPI-RMA optimized for thread scaling, which allows the code to run extremely efficiently at very high thread counts ideal for the latest generation of multi-core and many-core architectures. Such performance characteristics are needed in the era of “exascale” computing. We describe and demonstrate our high-performance design in detail with the intent that it may be used as a model for other, future astrophysical codes intended for applications demanding exceptional performance.

  10. Survey of computer codes applicable to waste facility performance evaluations

    International Nuclear Information System (INIS)

    Alsharif, M.; Pung, D.L.; Rivera, A.L.; Dole, L.R.

    1988-01-01

    This study is an effort to review existing information that is useful to develop an integrated model for predicting the performance of a radioactive waste facility. A summary description of 162 computer codes is given. The identified computer programs address the performance of waste packages, waste transport and equilibrium geochemistry, hydrological processes in unsaturated and saturated zones, and general waste facility performance assessment. Some programs also deal with thermal analysis, structural analysis, and special purposes. A number of these computer programs are being used by the US Department of Energy, the US Nuclear Regulatory Commission, and their contractors to analyze various aspects of waste package performance. Fifty-five of these codes were identified as being potentially useful on the analysis of low-level radioactive waste facilities located above the water table. The code summaries include authors, identification data, model types, and pertinent references. 14 refs., 5 tabs

  11. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  12. Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

  13. Towards a Framework for Generating Tests to Satisfy Complex Code Coverage in Java Pathfinder

    Science.gov (United States)

    Staats, Matt

    2009-01-01

    We present work on a prototype tool based on the JavaPathfinder (JPF) model checker for automatically generating tests satisfying the MC/DC code coverage criterion. Using the Eclipse IDE, developers and testers can quickly instrument Java source code with JPF annotations covering all MC/DC coverage obligations, and JPF can then be used to automatically generate tests that satisfy these obligations. The prototype extension to JPF enables various tasks useful in automatic test generation to be performed, such as test suite reduction and execution of generated tests.

  14. Benchmark testing and independent verification of the VS2DT computer code

    International Nuclear Information System (INIS)

    McCord, J.T.

    1994-11-01

    The finite difference flow and transport simulator VS2DT was benchmark tested against several other codes which solve the same equations (Richards equation for flow and the Advection-Dispersion equation for transport). The benchmark problems investigated transient two-dimensional flow in a heterogeneous soil profile with a localized water source at the ground surface. The VS2DT code performed as well as or better than all other codes when considering mass balance characteristics and computational speed. It was also rated highly relative to the other codes with regard to ease-of-use. Following the benchmark study, the code was verified against two analytical solutions, one for two-dimensional flow and one for two-dimensional transport. These independent verifications show reasonable agreement with the analytical solutions, and complement the one-dimensional verification problems published in the code's original documentation

  15. A good performance watermarking LDPC code used in high-speed optical fiber communication system

    Science.gov (United States)

    Zhang, Wenbo; Li, Chao; Zhang, Xiaoguang; Xi, Lixia; Tang, Xianfeng; He, Wenxue

    2015-07-01

    A watermarking LDPC code, which is a strategy designed to improve the performance of the traditional LDPC code, was introduced. By inserting some pre-defined watermarking bits into original LDPC code, we can obtain a more correct estimation about the noise level in the fiber channel. Then we use them to modify the probability distribution function (PDF) used in the initial process of belief propagation (BP) decoding algorithm. This algorithm was tested in a 128 Gb/s PDM-DQPSK optical communication system and results showed that the watermarking LDPC code had a better tolerances to polarization mode dispersion (PMD) and nonlinearity than that of traditional LDPC code. Also, by losing about 2.4% of redundancy for watermarking bits, the decoding efficiency of the watermarking LDPC code is about twice of the traditional one.

  16. Iterative optimization of performance libraries by hierarchical division of codes

    International Nuclear Information System (INIS)

    Donadio, S.

    2007-09-01

    The increasing complexity of hardware features incorporated in modern processors makes high performance code generation very challenging. Library generators such as ATLAS, FFTW and SPIRAL overcome this issue by empirically searching in the space of possible program versions for the one that performs the best. This thesis explores fully automatic solution to adapt a compute-intensive application to the target architecture. By mimicking complex sequences of transformations useful to optimize real codes, we show that generative programming is a practical tool to implement a new hierarchical compilation approach for the generation of high performance code relying on the use of state-of-the-art compilers. As opposed to ATLAS, this approach is not application-dependant but can be applied to fairly generic loop structures. Our approach relies on the decomposition of the original loop nest into simpler kernels. These kernels are much simpler to optimize and furthermore, using such codes makes the performance trade off problem much simpler to express and to solve. Finally, we propose a new approach for the generation of performance libraries based on this decomposition method. We show that our method generates high-performance libraries, in particular for BLAS. (author)

  17. Analysis of Phenix end-of-life natural convection test with the MARS-LMR code

    International Nuclear Information System (INIS)

    Jeong, H. Y.; Ha, K. S.; Lee, K. L.; Chang, W. P.; Kim, Y. I.

    2012-01-01

    The end-of-life test of Phenix reactor performed by the CEA provided an opportunity to have reliable and valuable test data for the validation and verification of a SFR system analysis code. KAERI joined this international program for the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main objectives of this study were to evaluate the capability of existing SFR system analysis code MARS-LMR and to identify any limitation of the code. The analysis was performed in three stages: pre-test analysis, blind posttest analysis, and final post-test analysis. In the pre-test analysis, the design conditions provided by the CEA were used to obtain a prediction of the test. The blind post-test analysis was based on the test conditions measured during the tests but the test results were not provided from the CEA. The final post-test analysis was performed to predict the test results as accurate as possible by improving the previous modeling of the test. Based on the pre-test analysis and blind test analysis, the modeling for heat structures in the hot pool and cold pool, steel structures in the core, heat loss from roof and vessel, and the flow path at core outlet were reinforced in the final analysis. The results of the final post-test analysis could be characterized into three different phases. In the early phase, the MARS-LMR simulated the heat-up process correctly due to the enhanced heat structure modeling. In the mid phase before the opening of SG casing, the code reproduced the decrease of core outlet temperature successfully. Finally, in the later phase the increase of heat removal by the opening of the SG opening was well predicted with the MARS-LMR code. (authors)

  18. Analysis of parallel computing performance of the code MCNP

    International Nuclear Information System (INIS)

    Wang Lei; Wang Kan; Yu Ganglin

    2006-01-01

    Parallel computing can reduce the running time of the code MCNP effectively. With the MPI message transmitting software, MCNP5 can achieve its parallel computing on PC cluster with Windows operating system. Parallel computing performance of MCNP is influenced by factors such as the type, the complexity level and the parameter configuration of the computing problem. This paper analyzes the parallel computing performance of MCNP regarding with these factors and gives measures to improve the MCNP parallel computing performance. (authors)

  19. Structure of fuel performance audit code for SFR metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

  20. Performance of FSO-OFDM based on BCH code

    Directory of Open Access Journals (Sweden)

    Jiao Xiao-lu

    2016-01-01

    Full Text Available As contrasted with the traditional OOK (on-off key system, FSO-OFDM system can resist the atmospheric scattering and improve the spectrum utilization rate effectively. Due to the instability of the atmospheric channel, the system will be affected by various factors, and resulting in a high BER. BCH code has a good error correcting ability, particularly in the short-length and medium-length code, and its performance is close to the theoretical value. It not only can check the burst errors but also can correct the random errors. Therefore, the BCH code is applied to the system to reduce the system BER. At last, the semi-physical simulation has been conducted with MATLAB. The simulation results show that when the BER is 10-2, the performance of OFDM is superior 4dB compared with OOK. In different weather conditions (extension rain, advection fog, dust days, when the BER is 10-5, the performance of BCH (255,191 channel coding is superior 4~5dB compared with uncoded system. All in all, OFDM technology and BCH code can reduce the system BER.

  1. Enhanced Verification Test Suite for Physics Simulation Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kamm, J R; Brock, J S; Brandon, S T; Cotrell, D L; Johnson, B; Knupp, P; Rider, W; Trucano, T; Weirs, V G

    2008-10-10

    This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations. The key points of this document are: (1) Verification deals with mathematical correctness of the numerical algorithms in a code, while validation deals with physical correctness of a simulation in a regime of interest. This document is about verification. (2) The current seven-problem Tri-Laboratory Verification Test Suite, which has been used for approximately five years at the DOE WP laboratories, is limited. (3) Both the methodology for and technology used in verification analysis have evolved and been improved since the original test suite was proposed. (4) The proposed test problems are in three basic areas: (a) Hydrodynamics; (b) Transport processes; and (c) Dynamic strength-of-materials. (5) For several of the proposed problems we provide a 'strong sense verification benchmark', consisting of (i) a clear mathematical statement of the problem with sufficient information to run a computer simulation, (ii) an explanation of how the code result and benchmark solution are to be evaluated, and (iii) a description of the acceptance criterion for simulation code results. (6) It is proposed that the set of verification test problems with which any particular code be evaluated include some of the problems described in this document. Analysis of the proposed verification test problems constitutes part of a necessary--but not sufficient--step that builds confidence in physics and engineering simulation codes. More complicated test cases, including physics models of

  2. Transient and fuel performance analysis with VTT's coupled code system

    International Nuclear Information System (INIS)

    Daavittila, A.; Hamalainen, A.; Raty, H.

    2005-01-01

    VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

  3. Severe fuel-damage scoping test performance

    International Nuclear Information System (INIS)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

  4. A development of containment performance analysis methodology using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-10-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.

  5. Fusion PIC code performance analysis on the Cori KNL system

    Energy Technology Data Exchange (ETDEWEB)

    Koskela, Tuomas S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Deslippe, Jack [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Friesen, Brian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Raman, Karthic [INTEL Corp. (United States)

    2017-05-25

    We study the attainable performance of Particle-In-Cell codes on the Cori KNL system by analyzing a miniature particle push application based on the fusion PIC code XGC1. We start from the most basic building blocks of a PIC code and build up the complexity to identify the kernels that cost the most in performance and focus optimization efforts there. Particle push kernels operate at high AI and are not likely to be memory bandwidth or even cache bandwidth bound on KNL. Therefore, we see only minor benefits from the high bandwidth memory available on KNL, and achieving good vectorization is shown to be the most beneficial optimization path with theoretical yield of up to 8x speedup on KNL. In practice we are able to obtain up to a 4x gain from vectorization due to limitations set by the data layout and memory latency.

  6. A development of containment performance analysis methodology using GOTHIC code

    International Nuclear Information System (INIS)

    Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.

    2003-01-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code

  7. SURE: a system of computer codes for performing sensitivity/uncertainty analyses with the RELAP code

    International Nuclear Information System (INIS)

    Bjerke, M.A.

    1983-02-01

    A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible

  8. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  9. Comparison of LIFE-4 and TEMECH code predictions with TREAT transient test data

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Bard, F.E.; Hunter, C.W.

    1984-09-01

    Transient tests in the TREAT reactor were performed on FFTF Reference design mixed-oxide fuel pins, most of which had received prior steady-state irradiation in the EBR-II reactor. These transient test results provide a data base for calibration and verification of fuel performance codes and for evaluation of processes that affect pin damage during transient events. This paper presents a comparison of the LIFE-4 and TEMECH fuel pin thermal/mechanical analysis codes with the results from 20 HEDL TREAT experiments, ten of which resulted in pin failure. Both the LIFE-4 and TEMECH codes provided an adequate representation of the thermal and mechanical data from the TREAT experiments. Also, a criterion for 50% probability of pin failure was developed for each code using an average cumulative damage fraction value calculated for the pins that failed. Both codes employ the two major cladding loading mechanisms of differential thermal expansion and central cavity pressurization which were demonstrated by the test results. However, a detailed evaluation of the code predictions shows that the two code systems weigh the loading mechanism differently to reach the same end points of the TREAT transient results

  10. Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code

    Science.gov (United States)

    Ruehl, K.; Michelen, C.; Gunawan, B.; Bosma, B.; Simmons, A.; Lomonaco, P.

    2015-12-01

    WEC-Sim is an open source code to model wave energy converters performance in operational waves, developed by Sandia and NREL and funded by the US DOE. The code is a time-domain modeling tool developed in MATLAB/SIMULINK using the multibody dynamics solver SimMechanics, and solves the WEC's governing equations of motion using the Cummins time-domain impulse response formulation in 6 degrees of freedom. The WEC-Sim code has undergone verification through code-to-code comparisons; however validation of the code has been limited to publicly available experimental data sets. While these data sets provide preliminary code validation, the experimental tests were not explicitly designed for code validation, and as a result are limited in their ability to validate the full functionality of the WEC-Sim code. Therefore, dedicated physical model tests for WEC-Sim validation have been performed. This presentation provides an overview of the WEC-Sim validation experimental wave tank tests performed at the Oregon State University's Directional Wave Basin at Hinsdale Wave Research Laboratory. Phase 1 of experimental testing was focused on device characterization and completed in Fall 2015. Phase 2 is focused on WEC performance and scheduled for Winter 2015/2016. These experimental tests were designed explicitly to validate the performance of WEC-Sim code, and its new feature additions. Upon completion, the WEC-Sim validation data set will be made publicly available to the wave energy community. For the physical model test, a controllable model of a floating wave energy converter has been designed and constructed. The instrumentation includes state-of-the-art devices to measure pressure fields, motions in 6 DOF, multi-axial load cells, torque transducers, position transducers, and encoders. The model also incorporates a fully programmable Power-Take-Off system which can be used to generate or absorb wave energy. Numerical simulations of the experiments using WEC-Sim will be

  11. Verification test calculations for the Source Term Code Package

    International Nuclear Information System (INIS)

    Denning, R.S.; Wooton, R.O.; Alexander, C.A.; Curtis, L.A.; Cybulskis, P.; Gieseke, J.A.; Jordan, H.; Lee, K.W.; Nicolosi, S.L.

    1986-07-01

    The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs

  12. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  13. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  14. Summary of aerosol code-comparison results for LWR aerosol containment tests LA1, LA2, and LA3

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1987-01-01

    The light-water reactor (LWR) aerosol containment experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, pretest calculations (for code-to-code comparisons) and blind post-test calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3

  15. Current Status of the LIFE Fast Reactors Fuel Performance Codes

    International Nuclear Information System (INIS)

    Yacout, A.M.; Billone, M.C.

    2013-01-01

    The LIFE-4 (Rev. 1) code was calibrated and validated using data from (U,Pu)O2 mixed-oxide fuel pins and UO2 blanket rods which were irradiation tested under steady-state and transient conditions. – It integrates a broad material and fuel-pin irradiation database into a consistent framework for use and extrapolation of the database to reactor design applications. – The code is available and running on different computer platforms (UNIX & PC) – Detailed documentations of the code’s models, routines, calibration and validation data sets are available. LIFE-METAL code is based on LIFE4 with modifications to include key phenomena applicable to metallic fuel, and metallic fuel properties – Calibrated with large database from irradiations in EBR-II – Further effort for calibration and detailed documentation. Recent activities with the codes are related to reactor design studies and support of licensing efforts for 4S and KAERI SFR designs. Future activities are related to re-assessment of the codes calibration and validation and inclusion of models for advanced fuels (transmutation fuels)

  16. Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation. Since 2009, IAEA has conducted a research program entitled as ICSP (International Collaborative Standard Problem) on integral PWR design to evaluate current the state of the art of thermal-hydraulic code in simulating natural circulation flow within integral type reactor. In this ICSP, experimental data obtained from MASLWR (Multi-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3)

  17. Performance Comparison of Assorted Color Spaces for Multilevel Block Truncation Coding based Face Recognition

    OpenAIRE

    H.B. Kekre; Sudeep Thepade; Karan Dhamejani; Sanchit Khandelwal; Adnan Azmi

    2012-01-01

    The paper presents a performance analysis of Multilevel Block Truncation Coding based Face Recognition among widely used color spaces. In [1], Multilevel Block Truncation Coding was applied on the RGB color space up to four levels for face recognition. Better results were obtained when the proposed technique was implemented using Kekre’s LUV (K’LUV) color space [25]. This was the motivation to test the proposed technique using assorted color spaces. For experimental analysis, two face databas...

  18. Comparison of computer code calculations with FEBA test data

    International Nuclear Information System (INIS)

    Zhu, Y.M.

    1988-06-01

    The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.) [de

  19. High performance computer code for molecular dynamics simulations

    International Nuclear Information System (INIS)

    Levay, I.; Toekesi, K.

    2007-01-01

    Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions

  20. Uniform peanut performance test 2017

    Science.gov (United States)

    The Uniform Peanut Performance Tests (UPPT) are designed to evaluate the commercial potential of advanced breeding peanut lines not formally released. The tests are performed in ten locations across the peanut production belt. In this study, 2 controls and 14 entries were evaluated at 8 locations....

  1. Inspection system performance test procedure

    International Nuclear Information System (INIS)

    Jensen, C.E.

    1995-01-01

    This procedure establishes requirements to administer a performance demonstration test. The test is to demonstrate that the double-shell tank inspection system (DSTIS) supplied by the contractor performs in accordance with the WHC-S-4108, Double-Shell Tank Ultrasonic Inspection Performance Specification, Rev. 2-A, January, 1995. The inspection system is intended to provide ultrasonic (UT) and visual data to determine integrity of the Westinghouse Hanford Company (WHC) site underground waste tanks. The robotic inspection system consists of the following major sub-systems (modules) and components: Mobile control center; Deployment module; Cable management assembly; Robot mechanism; Ultrasonic testing system; Visual testing system; Pneumatic system; Electrical system; and Control system

  2. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  3. Assessment of the system code DRUFAN/ATHLET using results of LOBI tests

    International Nuclear Information System (INIS)

    Burwell, J.M.; Kirmse, R.E.; Kyncl, M.; Malhotra, P.K.

    1989-09-01

    Four post-test analyses have been performed by GRS within the Shared Cost Action Programme (SCAP) sponsored by the Commission of the European Communities (contract 3015-86-07 EL ISP D) and by the Bundesminister fuer Forschung und Technologie of the Federal Republic of Germany (Research project RS 739). The four tests were mutually selected by the contractors (CEA, GRS, IKE, Univ. Pisa) of activity No. 3 and by the project organizer. Some of the tests were selected to be analyzed by more than one participant in order to allow comparison between analytical results obtained with different codes or obtained by different code-users. DRUFAN/ATHLET verification analyses were performed by IKE too. The four tests selected for the GRS activity are: - A2-77A (Natural Circulation Test), Analysis with ATHLET - A1-76 (Steam Generator Performance Test), Analysis with DRUFAN - BL-01 (Intermediate Leak), Analysis with ATHLET - A2-81 (Small Leak), Analysis with ATHLET. This final report contains the results of the four post test analysis including the comparison between measured and calculated quantities and the description of the applied codes, the selected model of the LOBI facility and the conclusions drawn for the improvement of the codes models

  4. Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

    International Nuclear Information System (INIS)

    Eom, Shin; Oh, Seung-Jong; Diab, Aya

    2018-01-01

    The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

  5. Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Shin; Oh, Seung-Jong; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering

    2018-02-15

    The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

  6. Performance of Low-Density Parity-Check Coded Modulation

    Science.gov (United States)

    Hamkins, Jon

    2010-01-01

    This paper reports the simulated performance of each of the nine accumulate-repeat-4-jagged-accumulate (AR4JA) low-density parity-check (LDPC) codes [3] when used in conjunction with binary phase-shift-keying (BPSK), quadrature PSK (QPSK), 8-PSK, 16-ary amplitude PSK (16- APSK), and 32-APSK.We also report the performance under various mappings of bits to modulation symbols, 16-APSK and 32-APSK ring scalings, log-likelihood ratio (LLR) approximations, and decoder variations. One of the simple and well-performing LLR approximations can be expressed in a general equation that applies to all of the modulation types.

  7. Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In this ICSP, experimental data obtained from MASLWR (Mulit-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are 1) loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels. In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation.

  8. Setting live coding performance in wider historical contexts

    OpenAIRE

    Norman, Sally Jane

    2016-01-01

    This paper sets live coding in the wider context of performing arts, construed as the poetic modelling and projection of liveness. Concepts of liveness are multiple, evolving, and scale-dependent: entities considered live from different cultural perspectives range from individual organisms and social groupings to entire ecosystems, and consequently reflect diverse temporal and spatial orders. Concepts of liveness moreover evolve with our tools, which generate and reveal new senses and places ...

  9. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    Van Uffelen, P.

    2015-01-01

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  10. Compressible Fluid Suspension Performance Testing

    National Research Council Canada - National Science Library

    Hoogterp, Francis

    2003-01-01

    ... compressible fluid suspension system that was designed and installed on the vehicle by DTI. The purpose of the tests was to evaluate the possible performance benefits of the compressible fluid suspension system...

  11. Where Lab Tests Are Performed

    Science.gov (United States)

    ... example, there may be sections that focus on microbiology, hematology, chemistry, and blood banking . Other units may perform highly specialized tests using electron microscopy and immunohistochemistry, and still others ...

  12. Using individual differences to test the role of temporal and place cues in coding frequency modulation.

    Science.gov (United States)

    Whiteford, Kelly L; Oxenham, Andrew J

    2015-11-01

    The question of how frequency is coded in the peripheral auditory system remains unresolved. Previous research has suggested that slow rates of frequency modulation (FM) of a low carrier frequency may be coded via phase-locked temporal information in the auditory nerve, whereas FM at higher rates and/or high carrier frequencies may be coded via a rate-place (tonotopic) code. This hypothesis was tested in a cohort of 100 young normal-hearing listeners by comparing individual sensitivity to slow-rate (1-Hz) and fast-rate (20-Hz) FM at a carrier frequency of 500 Hz with independent measures of phase-locking (using dynamic interaural time difference, ITD, discrimination), level coding (using amplitude modulation, AM, detection), and frequency selectivity (using forward-masking patterns). All FM and AM thresholds were highly correlated with each other. However, no evidence was obtained for stronger correlations between measures thought to reflect phase-locking (e.g., slow-rate FM and ITD sensitivity), or between measures thought to reflect tonotopic coding (fast-rate FM and forward-masking patterns). The results suggest that either psychoacoustic performance in young normal-hearing listeners is not limited by peripheral coding, or that similar peripheral mechanisms limit both high- and low-rate FM coding.

  13. Code accuracy evaluation of ISP 35 calculations based on NUPEC M-7-1 test

    International Nuclear Information System (INIS)

    Auria, F.D.; Oriolo, F.; Leonardi, M.; Paci, S.

    1995-01-01

    Quantitative evaluation of code uncertainties is a necessary step in the code assessment process, above all if best-estimate codes are utilised for licensing purposes. Aiming at quantifying the code accuracy, an integral methodology based on the Fast Fourier Transform (FFT) has been developed at the University of Pisa (DCMN) and has been already applied to several calculations related to primary system test analyses. This paper deals with the first application of the FFT based methodology to containment code calculations based on a hydrogen mixing and distribution test performed in the NUPEC (Nuclear Power Engineering Corporation) facility. It is referred to pre-test and post-test calculations submitted for the International Standard Problem (ISP) n. 35. This is a blind exercise, simulating the effects of steam injection and spray behaviour on gas distribution and mixing. The result of the application of this methodology to nineteen selected variables calculated by ten participants are here summarized, and the comparison (where possible) of the accuracy evaluated for the pre-test and for the post-test calculations of a same user is also presented. (author)

  14. ATES/heat pump simulations performed with ATESSS code

    Science.gov (United States)

    Vail, L. W.

    1989-01-01

    Modifications to the Aquifer Thermal Energy Storage System Simulator (ATESSS) allow simulation of aquifer thermal energy storage (ATES)/heat pump systems. The heat pump algorithm requires a coefficient of performance (COP) relationship of the form: COP = COP sub base + alpha (T sub ref minus T sub base). Initial applications of the modified ATES code to synthetic building load data for two sizes of buildings in two U.S. cities showed insignificant performance advantage of a series ATES heat pump system over a conventional groundwater heat pump system. The addition of algorithms for a cooling tower and solar array improved performance slightly. Small values of alpha in the COP relationship are the principal reason for the limited improvement in system performance. Future studies at Pacific Northwest Laboratory (PNL) are planned to investigate methods to increase system performance using alternative system configurations and operations scenarios.

  15. Performance testing With JMeter 29

    CERN Document Server

    Erinle, Bayo

    2013-01-01

    Performance Testing With JMeter 2.9 is a standard tutorial that will help you polish your fundamentals, guide you through various advanced topics, and along the process help you learn new tools and skills.This book is for developers, quality assurance engineers, testers, and test managers new to Apache JMeter, or those who are looking to get a good grounding in how to effectively use and become proficient with it. No prior testing experience is required.

  16. Performance analysis of LDPC codes on OOK terahertz wireless channels

    Science.gov (United States)

    Chun, Liu; Chang, Wang; Jun-Cheng, Cao

    2016-02-01

    Atmospheric absorption, scattering, and scintillation are the major causes to deteriorate the transmission quality of terahertz (THz) wireless communications. An error control coding scheme based on low density parity check (LDPC) codes with soft decision decoding algorithm is proposed to improve the bit-error-rate (BER) performance of an on-off keying (OOK) modulated THz signal through atmospheric channel. The THz wave propagation characteristics and channel model in atmosphere is set up. Numerical simulations validate the great performance of LDPC codes against the atmospheric fading and demonstrate the huge potential in future ultra-high speed beyond Gbps THz communications. Project supported by the National Key Basic Research Program of China (Grant No. 2014CB339803), the National High Technology Research and Development Program of China (Grant No. 2011AA010205), the National Natural Science Foundation of China (Grant Nos. 61131006, 61321492, and 61204135), the Major National Development Project of Scientific Instrument and Equipment (Grant No. 2011YQ150021), the National Science and Technology Major Project (Grant No. 2011ZX02707), the International Collaboration and Innovation Program on High Mobility Materials Engineering of the Chinese Academy of Sciences, and the Shanghai Municipal Commission of Science and Technology (Grant No. 14530711300).

  17. Generating performance portable geoscientific simulation code with Firedrake (Invited)

    Science.gov (United States)

    Ham, D. A.; Bercea, G.; Cotter, C. J.; Kelly, P. H.; Loriant, N.; Luporini, F.; McRae, A. T.; Mitchell, L.; Rathgeber, F.

    2013-12-01

    This presentation will demonstrate how a change in simulation programming paradigm can be exploited to deliver sophisticated simulation capability which is far easier to programme than are conventional models, is capable of exploiting different emerging parallel hardware, and is tailored to the specific needs of geoscientific simulation. Geoscientific simulation represents a grand challenge computational task: many of the largest computers in the world are tasked with this field, and the requirements of resolution and complexity of scientists in this field are far from being sated. However, single thread performance has stalled, even sometimes decreased, over the last decade, and has been replaced by ever more parallel systems: both as conventional multicore CPUs and in the emerging world of accelerators. At the same time, the needs of scientists to couple ever-more complex dynamics and parametrisations into their models makes the model development task vastly more complex. The conventional approach of writing code in low level languages such as Fortran or C/C++ and then hand-coding parallelism for different platforms by adding library calls and directives forces the intermingling of the numerical code with its implementation. This results in an almost impossible set of skill requirements for developers, who must simultaneously be domain science experts, numericists, software engineers and parallelisation specialists. Even more critically, it requires code to be essentially rewritten for each emerging hardware platform. Since new platforms are emerging constantly, and since code owners do not usually control the procurement of the supercomputers on which they must run, this represents an unsustainable development load. The Firedrake system, conversely, offers the developer the opportunity to write PDE discretisations in the high-level mathematical language UFL from the FEniCS project (http://fenicsproject.org). Non-PDE model components, such as parametrisations

  18. Independent validation testing of the FLAME computer code, Version 1.0

    International Nuclear Information System (INIS)

    Martian, P.; Chung, J.N.

    1992-07-01

    Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions

  19. Preliminary test conditions for KNGR SBLOCA DVI ECCS performance test

    International Nuclear Information System (INIS)

    Bae, Kyoo Whan; Song, Jin Ho; Chung, Young Jong; Sim, Suk Ku; Park, Jong Kyun

    1999-03-01

    The Korean Next Generation Reactor (KNGR) adopts 4-train Direct Vessel Injection (DVI) configuration and injects the safety injection water directly into the downcomer through the 8.5'' DVI nozzle. Thus, the thermal hydraulic phenomena such as ECCS mixing and bypass are expected to be different from those observed in the cold leg injection. In order to investigate the realistic injection phenomena and modify the analysis code developed in the basis of cold leg injection, thermal hydraulic test with the performance evaluation is required. Preliminarily, the sequence of events and major thermal hydraulic phenomena during the small break LOCA for KNGR are identified from the analysis results calculated by the CEFLASH-4AS/REM. It is shown from the analysis results that the major transient behaviors including the core mixture level are largely affected by the downcomer modeling. Therefore, to investigate the proper thermal hydraulic phenomena occurring in the downcomer with limited budget and time, the separate effects test focusing on this region is considered to be effective and the conceptual test facility based on this recommended. For this test facility the test initial and boundary conditions are developed using the CEFLASH-4AS/REM analysis results that will be used as input for the preliminary test requirements. The final test requirements will be developed through the further discussions with the test performance group. (Author). 10 refs., 18 tabs., 4 figs

  20. On the Performance of a Multi-Edge Type LDPC Code for Coded Modulation

    NARCIS (Netherlands)

    Cronie, H.S.

    2005-01-01

    We present a method to combine error-correction coding and spectral-efficient modulation for transmission over the Additive White Gaussian Noise (AWGN) channel. The code employs signal shaping which can provide a so-called shaping gain. The code belongs to the family of sparse graph codes for which

  1. Analysis of selected Halden overpressure tests using the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G., E-mail: grigori.khvostov@psi.ch [Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland); Wiesenack, W. [Institute for Energy Technology – OECD Halden Reactor Project, P.O. Box 173, N-1751 Halden (Norway)

    2016-12-15

    Highlights: • We analyse four Halden overpressure tests. • We determine a critical overpressure value for lift-off in a BWR fuel sample. • We show the role of bonding in over-pressurized rod behaviour. • We analytically quantify the degree of bonding via its impact on cladding elongation. • We hypothesize on an effect of circumferential cracks on thermal fuel response to overpressure. • We estimate a thermal effect of circumferential cracks based on interpretation of the data. - Abstract: Four Halden overpressure (lift-off) tests using samples with uranium dioxide fuel pre-irradiated in power reactors to a burnup of 60 MWd/kgU are analyzed. The FALCON code coupled to a mechanistic model, GRSW-A for fission gas release and gaseous-bubble swelling is used for the calculation. The advanced version of the FALCON code is shown to be applicable to best-estimate predictive analysis of overpressure tests using rods without, or weak pellet-cladding bonding, as well as scoping analysis of tests with fuels where stronger pellet-cladding bonding occurs. Significant effects of bonding and fuel cracking/relocation on the thermal and mechanical behaviour of highly over-pressurized rods are shown. The effect of bonding is particularly pronounced in the tests with the PWR samples. The present findings are basically consistent with an earlier analysis based on a direct interpretation of the experimental data. Additionally, in this paper, the specific effects are quantified based on the comparison of the data with the results of calculation. It is concluded that the identified effects are largely beyond the current traditional fuel-rod licensing analysis methods.

  2. Performance Testing of Cutting Fluids

    DEFF Research Database (Denmark)

    Belluco, Walter

    The importance of cutting fluid performance testing has increased with documentation requirements of new cutting fluid formulations based on more sustainable products, as well as cutting with minimum quantity of lubrication and dry cutting. Two sub-problems have to be solved: i) which machining...... tests feature repeatability, reproducibility and sensitivity to cutting fluids, and ii) to what extent results of one test ensure relevance to a wider set of machining situations. The present work is aimed at assessing the range of validity of the different testing methods, investigating correlation...... within the whole range of operations, materials, cutting fluids, operating conditions, etc. Cutting fluid performance was evaluated in turning, drilling, reaming and tapping, and with respect to tool life, cutting forces, chip formation and product quality (dimensional accuracy and surface integrity...

  3. COMPASS: A source term code for investigating capillary barrier performance

    International Nuclear Information System (INIS)

    Zhou, Wei; Apted, J.J.

    1996-01-01

    A computer code COMPASS based on compartment model approach is developed to calculate the near-field source term of the High-Level-Waste repository under unsaturated conditions. COMPASS is applied to evaluate the expected performance of Richard's (capillary) barriers as backfills to divert infiltrating groundwater at Yucca Mountain. Comparing the release rates of four typical nuclides with and without the Richard's barrier, it is shown that the Richard's barrier significantly decreases the peak release rates from the Engineered-Barrier-System (EBS) into the host rock

  4. Performance analysis of simultaneous dense coding protocol under decoherence

    Science.gov (United States)

    Huang, Zhiming; Zhang, Cai; Situ, Haozhen

    2017-09-01

    The simultaneous dense coding (SDC) protocol is useful in designing quantum protocols. We analyze the performance of the SDC protocol under the influence of noisy quantum channels. Six kinds of paradigmatic Markovian noise along with one kind of non-Markovian noise are considered. The joint success probability of both receivers and the success probabilities of one receiver are calculated for three different locking operators. Some interesting properties have been found, such as invariance and symmetry. Among the three locking operators we consider, the SWAP gate is most resistant to noise and results in the same success probabilities for both receivers.

  5. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  6. Radioactive material packaging performance testing

    International Nuclear Information System (INIS)

    Romano, T.; Cruse, J.M.

    1991-02-01

    To provide uniform packaging of hazardous materials on an international level, the United Nations has developed packaging recommendations that have been implemented worldwide. The United Nations packaging recommendations are performance oriented, allowing for a wide variety of package materials and systems. As a result of this international standard, efforts in the United States are being directed toward use of performance-oriented packaging and elimination of specification (designed) packaging. This presentation will focus on trends, design evaluation, and performance testing of radioactive material packaging. The impacts of US Department of Transportation Dockets HM-181 and HM-169A on specification and low-specific activity radioactive material packaging requirements are briefly discussed. The US Department of Energy's program for evaluating radioactive material packings per US Department of Transportation Specification 7A Type A requirements, is used as the basis for discussing low-activity packaging performance test requirements. High-activity package testing requirements are presented with examples of testing performed at the Hanford Site that is operated by Westinghouse Hanford Company for the US Department of Energy. 5 refs., 2 tabs

  7. Performance Analysis for Cooperative Communication System with QC-LDPC Codes Constructed with Integer Sequences

    Directory of Open Access Journals (Sweden)

    Yan Zhang

    2015-01-01

    Full Text Available This paper presents four different integer sequences to construct quasi-cyclic low-density parity-check (QC-LDPC codes with mathematical theory. The paper introduces the procedure of the coding principle and coding. Four different integer sequences constructing QC-LDPC code are compared with LDPC codes by using PEG algorithm, array codes, and the Mackey codes, respectively. Then, the integer sequence QC-LDPC codes are used in coded cooperative communication. Simulation results show that the integer sequence constructed QC-LDPC codes are effective, and overall performance is better than that of other types of LDPC codes in the coded cooperative communication. The performance of Dayan integer sequence constructed QC-LDPC is the most excellent performance.

  8. GEM: Performance and aging tests

    International Nuclear Information System (INIS)

    Cho, H.S.; Kadyk, J.; Han, S.H.; Hong, W.S.; Perez-Mendez, V.; Wenzel, W.; Pitts, K.; Martin, M.D.; Hutchins, J.B.

    1999-01-01

    Performance and aging tests have been done to characterize Gas Electron Multipliers (GEMs), including further design improvements such as a thicker GEM and a closed GEM. Since the effective GEM gain is typically smaller than the absolute GEM gain, due to trapping of avalanche electrons at the bottom GEM electrode, the authors performed field simulations and measurements for better understanding, and discuss methods to eliminate this effect. Other performance parameters of the GEMs are also presented, including absolute GEM gain, short-term and long-term gain stabilities

  9. Effect of two doses of ginkgo biloba extract (EGb 761) on the dual-coding test in elderly subjects.

    Science.gov (United States)

    Allain, H; Raoul, P; Lieury, A; LeCoz, F; Gandon, J M; d'Arbigny, P

    1993-01-01

    The subjects of this double-blind study were 18 elderly men and women (mean age, 69.3 years) with slight age-related memory impairment. In a crossover-study design, each subject received placebo or an extract of Ginkgo biloba (EGb 761) (320 mg or 600 mg) 1 hour before performing a dual-coding test that measures the speed of information processing; the test consists of several coding series of drawings and words presented at decreasing times of 1920, 960, 480, 240, and 120 ms. The dual-coding phenomenon (a break point between coding verbal material and images) was demonstrated in all the tests. After placebo, the break point was observed at 960 ms and dual coding beginning at 1920 ms. After each dose of the ginkgo extract, the break point (at 480 ms) and dual coding (at 960 ms) were significantly shifted toward a shorter presentation time, indicating an improvement in the speed of information processing.

  10. Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-07-15

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

  11. Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

    2012-07-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

  12. The error performance analysis over cyclic redundancy check codes

    Science.gov (United States)

    Yoon, Hee B.

    1991-06-01

    The burst error is generated in digital communication networks by various unpredictable conditions, which occur at high error rates, for short durations, and can impact services. To completely describe a burst error one has to know the bit pattern. This is impossible in practice on working systems. Therefore, under the memoryless binary symmetric channel (MBSC) assumptions, the performance evaluation or estimation schemes for digital signal 1 (DS1) transmission systems carrying live traffic is an interesting and important problem. This study will present some analytical methods, leading to efficient detecting algorithms of burst error using cyclic redundancy check (CRC) code. The definition of burst error is introduced using three different models. Among the three burst error models, the mathematical model is used in this study. The probability density function, function(b) of burst error of length b is proposed. The performance of CRC-n codes is evaluated and analyzed using function(b) through the use of a computer simulation model within CRC block burst error. The simulation result shows that the mean block burst error tends to approach the pattern of the burst error which random bit errors generate.

  13. CSNI Integral Test Facility Matrices for Validation of Best-Estimate Thermal-Hydraulic Computer Codes

    International Nuclear Information System (INIS)

    Glaeser, H.

    2008-01-01

    Internationally agreed Integral Test Facility (ITF) matrices for validation of realistic thermal hydraulic system computer codes were established. ITF development is mainly for Pressurised Water Reactors (PWRs) and Boiling Water Reactors (BWRs). A separate activity was for Russian Pressurised Water-cooled and Water-moderated Energy Reactors (WWER). Firstly, the main physical phenomena that occur during considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. In this paper some specific examples from the ITF matrices will also be provided. The matrices will be a guide for code validation, will be a basis for comparisons of code predictions performed with different system codes, and will contribute to the quantification of the uncertainty range of code model predictions. In addition to this objective, the construction of such a matrix is an attempt to record information which has been generated around the world over the last years, so that it is more accessible to present and future workers in that field than would otherwise be the case.

  14. Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

  15. Isotopic modelling using the ENIGMA-B fuel performance code

    International Nuclear Information System (INIS)

    Rossiter, G.D.; Cook, P.M.A.; Weston, R.

    2001-01-01

    A number of experimental programmes by BNFL and other MOX fabricators have now shown that the in-pile performance of MOX fuel is generally similar to that of conventional UO 2 fuel. Models based on UO 2 fuel experience form a good basis for a description of MOX fuel behaviour. However, an area where the performance of MOX fuel is sufficiently different from that of UO 2 to warrant model changes is in the radial power and burnup profile. The differences in radial power and burnup profile arise from the presence of significant concentrations of plutonium in MOX fuel, at beginning of life, and their subsequent evolution with burnup. Amongst other effects, plutonium has a greater neutron absorption cross-section than uranium. This paper focuses on the development of a new model for the radial power and burnup profile within a UO 2 or MOX fuel rod, in which the underlying fissile isotope concentration distributions are tracked during irradiation. The new model has been incorporated into the ENIGMA-B fuel performance code and has been extended to track the isotopic concentrations of the fission gases, xenon and krypton. The calculated distributions have been validated against results from rod puncture measurements and electron probe micro-analysis (EPMA) linescans, performed during the M501 post irradiation examination (PIE) programme. The predicted gas inventory of the fuel/clad gap is compared with the isotopic composition measured during rod puncture and the measured radial distributions of burnup (from neodymium measurements) and plutonium in the fuel are compared with the calculated distributions. It is shown that there is good agreement between the code predictions and the measurements. (author)

  16. Performance of the dot product function in radiative transfer code SORD

    Science.gov (United States)

    Korkin, Sergey; Lyapustin, Alexei; Sinyuk, Aliaksandr; Holben, Brent

    2016-10-01

    The successive orders of scattering radiative transfer (RT) codes frequently call the scalar (dot) product function. In this paper, we study performance of some implementations of the dot product in the RT code SORD using 50 scenarios for light scattering in the atmosphere-surface system. In the dot product function, we use the unrolled loops technique with different unrolling factor. We also considered the intrinsic Fortran functions. We show results for two machines: ifort compiler under Windows, and pgf90 under Linux. Intrinsic DOT_PRODUCT function showed best performance for the ifort. For the pgf90, the dot product implemented with unrolling factor 4 was the fastest. The RT code SORD together with the interface that runs all the mentioned tests are publicly available from ftp://maiac.gsfc.nasa.gov/pub/skorkin/SORD_IP_16B (current release) or by email request from the corresponding (first) author.

  17. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  18. SCANAIR a transient fuel performance code Part two: Assessment of modelling capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Georgenthum, Vincent, E-mail: vincent.georgenthum@irsn.fr; Moal, Alain; Marchand, Olivier

    2014-12-15

    Highlights: • The SCANAIR code is devoted to the study of irradiated fuel rod behaviour during RIA. • The paper deals with the status of the code validation for PWR rods. • During the PCMI stage there is a good agreement between calculations and experiments. • The boiling crisis occurrence is rather well predicted. • The code assessment during the boiling crisis has still to be improved. - Abstract: In the frame of their research programmes on fuel safety, the French Institut de Radioprotection et de Sûreté Nucléaire develops the SCANAIR code devoted to the study of irradiated fuel rod behaviour during reactivity initiated accident. A first paper was focused on detailed modellings and code description. This second paper deals with the status of the code validation for pressurised water reactor rods performed thanks to the available experimental results. About 60 integral tests carried out in CABRI and NSRR experimental reactors and 24 separated tests performed in the PATRICIA facility (devoted to the thermal-hydraulics study) have been recalculated and compared to experimental data. During the first stage of the transient, the pellet clad mechanical interaction phase, there is a good agreement between calculations and experiments: the clad residual elongation and hoop strain of non failed tests but also the failure occurrence and failure enthalpy of failed tests are correctly calculated. After this first stage, the increase of cladding temperature can lead to the Departure from Nucleate Boiling. During the film boiling regime, the clad temperature can reach a very high temperature (>700 °C). If the boiling crisis occurrence is rather well predicted, the calculation of the clad temperature and the clad hoop strain during this stage have still to be improved.

  19. Water evaporation over sump surface in nuclear containment studies: CFD and LP codes validation on TOSQAN tests

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Degrees du Lou, O. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Arts et Métiers ParisTech, DynFluid Lab. EA92, 151, boulevard de l’Hôpital, 75013 Paris (France); Gelain, T. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France)

    2013-10-15

    Highlights: • Simulations of evaporative TOSQAN sump tests are performed. • These tests are under air–steam gas conditions with addition of He, CO{sub 2} and SF{sub 6}. • ASTEC-CPA LP and TONUS-CFD codes with UDF for sump model are used. • Validation of sump models of both codes show good results. • The code–experiment differences are attributed to turbulent gas mixing modeling. -- Abstract: During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The seven tests are air–steam tests, as well as tests with other non-condensable gases (He, CO{sub 2} and SF{sub 6}) under steady and transient conditions (two depressurization tests). The results show a good agreement between codes and experiments, indicating a good behavior of the sump models in both codes. The sump model developed as User-Defined Functions (UDF) for TONUS is considered as well validated and is ‘ready-to-use’ for all CFD codes in which such UDF can be added. The remaining discrepancies between codes and experiments are caused by turbulent transport and gas mixing, especially in the presence of non-condensable gases other than air, so that code validation on this important topic for hydrogen safety analysis is still recommended.

  20. A fast and compact Fuel Rod Performance Simulator code for predictive, interpretive and educational purpose

    International Nuclear Information System (INIS)

    Lorenzen, J.

    1990-01-01

    A new Fuel rod Performance Simulator code FRPS has been developed, tested and benchmarked and is now available in different versions. The user may choose between the batch version INTERPIN producing results in form of listings or beforehand defined plots, or the interactive simulator code SIMSIM which is stepping through a power history under the control of user. Both versions are presently running on minicomputers and PC:s using EGA-Graphics. A third version is the implementation in a Studsvik Compact Simulator with FRPS being one of its various modules receiving the dynamic inputs from the simulator

  1. Software Design Document for the AMP Nuclear Fuel Performance Code

    International Nuclear Information System (INIS)

    Philip, Bobby; Clarno, Kevin T.; Cochran, Bill

    2010-01-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  2. Test of Effective Solid Angle code for the efficiency calculation of volume source

    Energy Technology Data Exchange (ETDEWEB)

    Kang, M. Y.; Kim, J. H.; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of); Sun, G. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is hard to determine a full energy (FE) absorption peak efficiency curve for an arbitrary volume source by experiment. That's why the simulation and semi-empirical methods have been preferred so far, and many works have progressed in various ways. Moens et al. determined the concept of effective solid angle by considering an attenuation effect of γ-rays in source, media and detector. This concept is based on a semi-empirical method. An Effective Solid Angle code (ESA code) has been developed for years by the Applied Nuclear Physics Group in Seoul National University. ESA code converts an experimental FE efficiency curve determined by using a standard point source to that for a volume source. To test the performance of ESA Code, we measured the point standard sources and voluminous certified reference material (CRM) sources of γ-ray, and compared with efficiency curves obtained in this study. 200∼1500 KeV energy region is fitted well. NIST X-ray mass attenuation coefficient data is used currently to check for the effect of linear attenuation only. We will use the interaction cross-section data obtained from XCOM code to check the each contributing factor like photoelectric effect, incoherent scattering and coherent scattering in the future. In order to minimize the calculation time and code simplification, optimization of algorithm is needed.

  3. Improvement and test calculation on basic code or sodium-water reaction jet

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Yoshinori; Itooka, Satoshi [Advanced Reactor Engineering Center, Hitachi Works, Hitachi Ltd., Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo [Consulting Engineering Dept., Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

    1999-03-01

    In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3{center_dot}Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

  4. Improvement and test calculation on basic code or sodium-water reaction jet

    International Nuclear Information System (INIS)

    Saito, Yoshinori; Itooka, Satoshi; Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo

    1999-03-01

    In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

  5. BER performance comparison of optical CDMA systems with/without turbo codes

    Science.gov (United States)

    Kulkarni, Muralidhar; Chauhan, Vijender S.; Dutta, Yashpal; Sinha, Ravindra K.

    2002-08-01

    In this paper, we have analyzed and simulated the BER performance of a turbo coded optical code-division multiple-access (TC-OCDMA) system. A performance comparison has been made between uncoded OCDMA and TC-OCDMA systems employing various OCDMA address codes (optical orthogonal codes (OOCs), Generalized Multiwavelength Prime codes (GMWPC's), and Generalized Multiwavelength Reed Solomon code (GMWRSC's)). The BER performance of TC-OCDMA systems has been analyzed and simulated by varying the code weight of address code employed by the system. From the simulation results, it is observed that lower weight address codes can be employed for TC-OCDMA systems that can have the equivalent BER performance of uncoded systems employing higher weight address codes for a fixed number of active users.

  6. Early Experiences Writing Performance Portable OpenMP 4 Codes

    Energy Technology Data Exchange (ETDEWEB)

    Joubert, Wayne [ORNL; Hernandez, Oscar R [ORNL

    2016-01-01

    In this paper, we evaluate the recently available directives in OpenMP 4 to parallelize a computational kernel using both the traditional shared memory approach and the newer accelerator targeting capabilities. In addition, we explore various transformations that attempt to increase application performance portability, and examine the expressiveness and performance implications of using these approaches. For example, we want to understand if the target map directives in OpenMP 4 improve data locality when mapped to a shared memory system, as opposed to the traditional first touch policy approach in traditional OpenMP. To that end, we use recent Cray and Intel compilers to measure the performance variations of a simple application kernel when executed on the OLCF s Titan supercomputer with NVIDIA GPUs and the Beacon system with Intel Xeon Phi accelerators attached. To better understand these trade-offs, we compare our results from traditional OpenMP shared memory implementations to the newer accelerator programming model when it is used to target both the CPU and an attached heterogeneous device. We believe the results and lessons learned as presented in this paper will be useful to the larger user community by providing guidelines that can assist programmers in the development of performance portable code.

  7. Fire-safety engineering and performance-based codes

    DEFF Research Database (Denmark)

    Sørensen, Lars Schiøtt

    project administrators, etc. The book deals with the following topics: • Historical presentation on the subject of fire • Legislation and building project administration • European fire standardization • Passive and active fire protection • Performance-based Codes • Fire-safety Engineering • Fundamental......Fire-safety Engineering is written as a textbook for Engineering students at universities and other institutions of higher education that teach in the area of fire. The book can also be used as a work of reference for consulting engineers, Building product manufacturers, contractors, building...... thermodynamics • Heat exchange during the fire process • Skin burns • Burning rate, energy release rate and design fires • Proposal to Risk-based design fires • Proposal to a Fire scale • Material ignition and flame spread • Fire dynamics in buildings • Combustion products and toxic gases • Smoke inhalation...

  8. Assessment of MARMOT. A Mesoscale Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Tonks, M. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fromm, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, M. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

  9. Development and validation of the ENIGMA code for MOX fuel performance modelling

    International Nuclear Information System (INIS)

    Palmer, I.; Rossiter, G.; White, R.J.

    2000-01-01

    The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

  10. Post-test analysis for the APR1400 LBLOCA DVI performance test using MARS

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Lee, Y. J.; Kim, H. C.; Bae, Y. Y.; Park, J. K.; Lee, W.

    2002-03-01

    Post-test analyses using a multi-dimensional best-estimate analysis code, MARS, are performed for the APR1400 LBLOCA DVI (Direct Vessel Injection) performance tests. This report describes the code evaluation results for the test data of various void height tests and direct bypass tests that have been performed at MIDAS test facility. MIDAS is a scaled test facility of APR1400 with the objective of identifying multi-dimensional thermal-hydraulic phenomena in the downcomer during the reflood conditions of a large break LOCA. A modified linear scale ratio was applied in its construction and test conditions. The major thermal-hydraulic parameters such as ECC bypass fraction, steam condensation fraction, and temperature distributions in downcomer are compared and evaluated. The evaluation results of MARS code for the various test cases show that: (a) MARS code has an advanced modeling capability of well predicting major multi-dimensional thermal-hydraulic phenomena occurring in the downcomer, (b) MARS code under-predicts the steam condensation rates, which in turn causes to over-predict the ECC bypass rates. However, the trend of decrease in steam condensation rate and increase in ECC bypass rate in accordance with the increase in steam flow rate, and the calculation results of the ECC bypass rates under the EM analysis conditions generally agree with the test data

  11. Sensitivity analysis on the interfacial drag in SPACE code to simulate UPTF separate effect test about loop seal clearance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sukho; Lim, Sanggyu; You, Gukjong; Park, Youngsheop [Korea Hydro and Nuclear Power Company, Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear thermal hydraulic system code known as SPACE (Safety and Performance Analysis CodE) was developed and its V and V (Verification and Validation) have been conducted using well-known SETs (Separate Effect Tests) and IETs (Integral Effect Tests). At the same time, the SBLOCA (Small Break Loss of Coolant Accident) methodology in accordance with Appendix K of 10CFR50 for the APR1400 (Advanced Power Reactor 1400) was developed and applied to regulatory body for licensing in 2013. Especially, the SBLOCA methodology developed using SPACE v2.14 code adopts inherent test matrix independent of V and V test to show its conservatism for important phenomena. In this paper, the predictability of SPACE code for UPTF (Upper Plenum Test Facility) test simulating loop seal clearance of SBLOCA important phenomena and the related sensitivity analysis are introduced.

  12. Performance Analysis of New Binary User Codes for DS-CDMA Communication

    Science.gov (United States)

    Usha, Kamle; Jaya Sankar, Kottareddygari

    2016-03-01

    This paper analyzes new binary spreading codes through correlation properties and also presents their performance over additive white Gaussian noise (AWGN) channel. The proposed codes are constructed using gray and inverse gray codes. In this paper, a n-bit gray code appended by its n-bit inverse gray code to construct the 2n-length binary user codes are discussed. Like Walsh codes, these binary user codes are available in sizes of power of two and additionally code sets of length 6 and their even multiples are also available. The simple construction technique and generation of code sets of different sizes are the salient features of the proposed codes. Walsh codes and gold codes are considered for comparison in this paper as these are popularly used for synchronous and asynchronous multi user communications respectively. In the current work the auto and cross correlation properties of the proposed codes are compared with those of Walsh codes and gold codes. Performance of the proposed binary user codes for both synchronous and asynchronous direct sequence CDMA communication over AWGN channel is also discussed in this paper. The proposed binary user codes are found to be suitable for both synchronous and asynchronous DS-CDMA communication.

  13. Development of a computer program to support an efficient non-regression test of a thermal-hydraulic system code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jun Yeob; Jeong, Jae Jun [School of Mechanical Engineering, Pusan National University, Busan (Korea, Republic of); Suh, Jae Seung [System Engineering and Technology Co., Daejeon (Korea, Republic of); Kim, Kyung Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    During the development process of a thermal-hydraulic system code, a non-regression test (NRT) must be performed repeatedly in order to prevent software regression. The NRT process, however, is time-consuming and labor-intensive. Thus, automation of this process is an ideal solution. In this study, we have developed a program to support an efficient NRT for the SPACE code and demonstrated its usability. This results in a high degree of efficiency for code development. The program was developed using the Visual Basic for Applications and designed so that it can be easily customized for the NRT of other computer codes.

  14. Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code

    International Nuclear Information System (INIS)

    Skerlavaj, A.; Mavko, B.; Kljenak, I.

    2001-01-01

    The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)

  15. Simulation of the KAERI PASCAL Test with MARS-KS and TRACE Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Won; Cheong, Aeju; Shin, Andong; Cho, Min Ki [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In order to validate the operational performance of the PAFS, KAERI has performed the experimental investigation using the PASCAL (PAFS Condensing heat removal Assessment Loop) facility. In this study, we simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. We simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. The calculated results of heat flux, inner wall surface temperature of the condensing tube, fluid temperature, and steam mass flow rate are compared with the experimental data. The result shows that the MARS-KS generally under-predict the heat fluxes. The TRACE over-predicts the heat flux at tube inlet region and under-predicts it at tube outlet region. The TRACE prediction shows larger amount of steam condensation by about 3% than the MARS-KS prediction.

  16. Radioactive material packaging performance testing

    International Nuclear Information System (INIS)

    Romano, T.

    1992-06-01

    In an effort to provide uniform packaging of hazardous material on an international level, recommendations for the transport of dangerous goods have been developed by the United Nations. These recommendations are performance oriented and contrast with a large number of packaging specifications in the US Department of Transportation's hazard materials regulations. This dual system presents problems when international shipments enter the US Department of Transportation's system. Faced with the question of continuing a dual system or aligning with the international system, the Research and Special Programs Administration of the US Department of Transportation responded with Docket HM-181. This began the transition toward the international transportation system. Following close behind is Docket HM-169A, which addressed low specific activity radioactive material packaging. This paper will discuss the differences between performance-oriented and specification packaging, the transition toward performance-oriented packaging by the US Department of Transportation, and performance-oriented testing of radioactive material packaging by Westinghouse Hanford Company. Dockets HM-181 and HM-169A will be discussed along with Type A (low activity) and Type B (high activity) radioactive material packaging evaluations

  17. Code Assessment of SPACE 2.19 using LSTF Steam Generator Tube Rupture Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The SPACE is a best estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. The present work is on the line of expanding the work by Kim et al. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run LSTF SB-SG-06 experiment simulating a Steam Generator Tube Rupture (SGTR) transient. In order to identify the predictability of SPACE 2.19, the LSTF steam generator tube rupture test was simulated. To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR and the RELAP5/ MOD3.1 are used. The calculation results indicate that the SPACE 2.19 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator relief valve.

  18. Knowledge and Performance about Nursing Ethic Codes from Nurses' and Patients' Perspective in Tabriz Teaching Hospitals, Iran.

    Science.gov (United States)

    Mohajjel-Aghdam, Alireza; Hassankhani, Hadi; Zamanzadeh, Vahid; Khameneh, Saied; Moghaddam, Sara

    2013-09-01

    Nursing profession requires knowledge of ethics to guide performance. The nature of this profession necessitates ethical care more than routine care. Today, worldwide definition of professional ethic code has been done based on human and ethical issues in the communication between nurse and patient. To improve all dimensions of nursing, we need to respect ethic codes. The aim of this study is to assess knowledge and performance about nursing ethic codes from nurses' and patients' perspective. A descriptive study Conducted upon 345 nurses and 500 inpatients in six teaching hospitals of Tabriz, 2012. To investigate nurses' knowledge and performance, data were collected by using structured questionnaires. Statistical analysis was done using descriptive and analytic statistics, independent t-test and ANOVA and Pearson correlation coefficient, in SPSS13. Most of the nurses were female, married, educated at BS degree and 86.4% of them were aware of Ethic codes also 91.9% of nurses and 41.8% of patients represented nurses respect ethic codes. Nurses' and patients' perspective about ethic codes differed significantly. Significant relationship was found between nurses' knowledge of ethic codes and job satisfaction and complaint of ethical performance. According to the results, consideration to teaching ethic codes in nursing curriculum for student and continuous education for staff is proposed, on the other hand recognizing failures of the health system, optimizing nursing care, attempt to inform patients about Nursing ethic codes, promote patient rights and achieve patient satisfaction can minimize the differences between the two perspectives.

  19. GetData Digitizing Program Code: Description, Testing, Training

    International Nuclear Information System (INIS)

    Taova, S.

    2013-01-01

    90 percents of compilation in our center is obtained by data digitizing. So we are rather interested in the development of different techniques of data digitizing. Plots containing a great amount of points and solid lines are most complicated for digitizing. From our point of view including to the Exfor-Digitizer procedures of automatic or semi-automatic digitizing will allow to simplify significantly this process. We managed to test some free available program codes. Program GETDATA Graph Digitizer (www.getdata- graph-digitizer.com) looks more suitable for our purposes. GetData Graph Digitizer is a program for digitizing graphs, plots and maps. Main features of GetData Graph Digitizer are: - supported graphics formats are TIFF, JPEG, BMP and PCX; - two algorithms for automatic digitizing; - convenient manual digitizing; - reorder tool for easy points reordering; - save/open workspace, which allows to save the work and return to it later; - obtained data can be exported to the clipboard; - export to the formats: TXT (text file), XLS (MS Excel), XML, DXF (AutoCAD) and EPS (PostScript). GetData Graph Digitizer includes two algorithms for automatic digitizing. Auto trace lines: This method is designed to digitize solid lines. Choose the starting point, and the program will trace the line, stopping at it's end. To trace the line use Operations =>Auto trace lines menu or context menu ('Auto trace lines' item). To choose starting point click left mouse button, or click right mouse button to additionally choose direction for line tracing. Digitize area: The second way is to set digitizing area. This method works for any type of lines, including dashed lines. Data points are set at the intersection of grid with the line. You can choose the type of grid (X grid or Y grid), and set the distance between grid lines. You can also make the grid be shifted in such a way, that it will pass through a specific X (or Y) value. To digitize area use Operations →Digitize area menu

  20. Impact of intra-flow network coding on the relay channel performance: an analytical study

    OpenAIRE

    Apavatjrut , Anya; Goursaud , Claire; Jaffrès-Runser , Katia; Gorce , Jean-Marie

    2012-01-01

    International audience; One of the most powerful ways to achieve trans- mission reliability over wireless links is to employ efficient coding techniques. This paper investigates the performance of a transmission over a relay channel where information is protected by two layers of coding. In the first layer, transmission reliability is ensured by fountain coding at the source. The second layer incorporates network coding at the relay node. Thus, fountain coded packets are re-encoded at the relay...

  1. Test Performance Related Dysfunctional Beliefs

    Directory of Open Access Journals (Sweden)

    Recep TÜTÜNCÜ

    2012-11-01

    Full Text Available Objective: Examinations by using tests are very frequently used in educational settings and successful studying before the examinations is a complex matter to deal with. In order to understand the determinants of success in exams better, we need to take into account not only emotional and motivational, but also cognitive aspects of the participants such as dysfunctional beliefs. Our aim is to present the relationship between candidates’ characteristics and distorted beliefs/schemata just before an examination. Method: The subjects of the study were 30 female and 30 male physicians who were about to take the medical specialization exam (MSE in Turkey. Dysfunctional Attitude Scale (DAS and Young Schema Questionnaire Short Form (YSQ-SF were applied to the subjects. The statistical analysis was done using the F test, Mann-Whitney, Kruskal-Wallis, chi-square test and spearman’s correlation test. Results: It was shown that some of the DAS and YSQ-SF scores were significantly higher in female gender, in the group who could not pass the exam, who had repetitive examinations, who had their first try taking an examination and who were unemployed at the time of the examination. Conclusion: Our findings indicate that candidates seeking help before MSE examination could be referred for cognitive therapy or counseling even they do not have any psychiatric diagnosis due to clinically significant cognitive distortion. Measurement and treatment of cognitive distortions that have negative impact on MSE performance may improve the cost-effectiveness and mental well being of the young doctors.

  2. Performance Evaluation of HARQ Technique with UMTS Turbo Code

    Directory of Open Access Journals (Sweden)

    S. S. Brkić

    2011-11-01

    Full Text Available The hybrid automatic repeat request technique (HARQ represents the error control principle which combines an error correcting code and automatic repeat request procedure (ARQ, within the same transmission system. In this paper, using Monte Carlo simulation process, the characteristics of HARQ technique are determined, for the case of the Universal Mobile Telecommunication System (UMTS turbo code.

  3. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Glass, R.E.

    1989-01-01

    Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes (''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Cask,'' R.E. Glass, Sandia National Laboratories, 1985; ''Sample Problem Manual for Benchmarking of Cask Analysis Codes,'' R.E. Glass, Sandia National Laboratories, 1988; ''Standard Thermal Problem Set for the Evaluation of Heat Transfer Codes Used in the Assessment of Transportation Packages, R.E. Glass, et al., Sandia National Laboratories, 1988) used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in ''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks,'' R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem. 6 refs., 5 figs

  4. Environmental performance of green building code and certification systems.

    Science.gov (United States)

    Suh, Sangwon; Tomar, Shivira; Leighton, Matthew; Kneifel, Joshua

    2014-01-01

    We examined the potential life-cycle environmental impact reduction of three green building code and certification (GBCC) systems: LEED, ASHRAE 189.1, and IgCC. A recently completed whole-building life cycle assessment (LCA) database of NIST was applied to a prototype building model specification by NREL. TRACI 2.0 of EPA was used for life cycle impact assessment (LCIA). The results showed that the baseline building model generates about 18 thousand metric tons CO2-equiv. of greenhouse gases (GHGs) and consumes 6 terajoule (TJ) of primary energy and 328 million liter of water over its life-cycle. Overall, GBCC-compliant building models generated 0% to 25% less environmental impacts than the baseline case (average 14% reduction). The largest reductions were associated with acidification (25%), human health-respiratory (24%), and global warming (GW) (22%), while no reductions were observed for ozone layer depletion (OD) and land use (LU). The performances of the three GBCC-compliant building models measured in life-cycle impact reduction were comparable. A sensitivity analysis showed that the comparative results were reasonably robust, although some results were relatively sensitive to the behavioral parameters, including employee transportation and purchased electricity during the occupancy phase (average sensitivity coefficients 0.26-0.29).

  5. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  6. Analyses of CsI aerosol deposition tests in WIND project with ART and VICTORIA codes

    International Nuclear Information System (INIS)

    Yuchi, Y.; Shibazaki, H.; Kudo, T.

    2000-01-01

    Deposition behavior of cesium iodide (CsI) was analyzed with ART and VICTORIA-92 codes for a test of the aerosol re-vaporization test series performed in WIND project at JAERI. In the test analyzed, CsI aerosol was injected into piping of test section where metaboric acid (HBO 2 ) was placed in advance on the floor area. It was confirmed in the present analysis that similar results on the CsI deposition were obtained between ART and VICTORIA when influences of chemical interactions were negligibly small. The analysis with VICTORIA agreed satisfactorily with the test results in analytical cases that cesium metaborate (CsBO 2 ) was injected into the test section instead of CsI to simulate the pre-existence of HBO 2 effect. (author)

  7. Results of aerosol code comparisons with releases from ACE MCCI tests

    International Nuclear Information System (INIS)

    Fink, J.K.; Corradini, M.; Hidaka, A.; Hontanon, E.; Mignanelli, M.A.; Schroedl, E.; Strizhov, V.

    1992-01-01

    Results of aerosol release calculations by six groups from six countries are compared with the releases from ACE MCCI Test L6. The codes used for these calculations included: SOLGASMIX-PV, SOLGASMIX Reactor 1986, CORCON.UW, VANESA 1.01, and CORCON mod2.04/VANESA 1.01. Calculations were performed with the standard VANESA 1.01 code and with modifications to the VANESA code such as the inclusion of various zirconium-silica chemical reactions. Comparisons of results from these calculations were made with Test L6 release fractions for U, Zr, Si, the fission-product elements Te, Ba, Sr, Ce, La, Mo and control materials Ag, In, and Ru. Reasonable agreement was obtained between calculations and Test L6 results for the volatile elements Ag, In and Te. Calculated releases of the low volatility fission products ranged from within an order of magnitude to five orders of magnitude of Test L6 values. Releases were over and underestimated by calculations. Poorest agreements were obtained for Mo and Si

  8. VINE-A NUMERICAL CODE FOR SIMULATING ASTROPHYSICAL SYSTEMS USING PARTICLES. II. IMPLEMENTATION AND PERFORMANCE CHARACTERISTICS

    International Nuclear Information System (INIS)

    Nelson, Andrew F.; Wetzstein, M.; Naab, T.

    2009-01-01

    We continue our presentation of VINE. In this paper, we begin with a description of relevant architectural properties of the serial and shared memory parallel computers on which VINE is intended to run, and describe their influences on the design of the code itself. We continue with a detailed description of a number of optimizations made to the layout of the particle data in memory and to our implementation of a binary tree used to access that data for use in gravitational force calculations and searches for smoothed particle hydrodynamics (SPH) neighbor particles. We describe the modifications to the code necessary to obtain forces efficiently from special purpose 'GRAPE' hardware, the interfaces required to allow transparent substitution of those forces in the code instead of those obtained from the tree, and the modifications necessary to use both tree and GRAPE together as a fused GRAPE/tree combination. We conclude with an extensive series of performance tests, which demonstrate that the code can be run efficiently and without modification in serial on small workstations or in parallel using the OpenMP compiler directives on large-scale, shared memory parallel machines. We analyze the effects of the code optimizations and estimate that they improve its overall performance by more than an order of magnitude over that obtained by many other tree codes. Scaled parallel performance of the gravity and SPH calculations, together the most costly components of most simulations, is nearly linear up to at least 120 processors on moderate sized test problems using the Origin 3000 architecture, and to the maximum machine sizes available to us on several other architectures. At similar accuracy, performance of VINE, used in GRAPE-tree mode, is approximately a factor 2 slower than that of VINE, used in host-only mode. Further optimizations of the GRAPE/host communications could improve the speed by as much as a factor of 3, but have not yet been implemented in VINE

  9. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Glass, R.E.

    1989-01-01

    Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks, R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem

  10. Performance Analysis of Wavelength Multiplexed Sac Ocdma Codes in Beat Noise Mitigation in Sac Ocdma Systems

    Science.gov (United States)

    Alhassan, A. M.; Badruddin, N.; Saad, N. M.; Aljunid, S. A.

    2013-07-01

    In this paper we investigate the use of wavelength multiplexed spectral amplitude coding (WM SAC) codes in beat noise mitigation in coherent source SAC OCDMA systems. A WM SAC code is a low weight SAC code, where the whole code structure is repeated diagonally (once or more) in the wavelength domain to achieve the same cardinality as a higher weight SAC code. Results show that for highly populated networks, the WM SAC codes provide better performance than SAC codes. However, for small number of active users the situation is reversed. Apart from their promising improvement in performance, these codes are more flexible and impose less complexity on the system design than their SAC counterparts.

  11. First vapor explosion calculations performed with MC3D thermal-hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires

    1998-01-01

    This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)

  12. Testing the new stochastic neutronic code ANET in simulating safety important parameters

    International Nuclear Information System (INIS)

    Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.

    2017-01-01

    Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement

  13. The application of RCM to ASME code requirements for in-service testing

    International Nuclear Information System (INIS)

    Rowley, C.W.

    1990-01-01

    This paper reports that the high reliability of nuclear power plant systems and components is highly important for both nuclear safety and electrical power production economics. The optimum operating performance of these plant systems and components is heavily dependent on the original or modified design for its inherent reliability and the appropriate trade-off in preventive and corrective maintenance for its developed reliability. In developing this optimum operating performance goal, the plant staff could rely solely on the experience of its personnel. However using this internal subjective approach, the average nuclear power availability has been far less than 80%. Obviously the production economics of a nuclear power plant is the province of the owner-operator, but the safety system and component performance impacts the entire industry. Hence the nuclear industry needs to have in-service testing requirements that maintain the necessary safety standards. Historically the ASME Inservice Testing Code has been a vehicle for defining some of those necessary safety standards, such as inservice testing of pumps, valves, and snubbers. The nuclear industry needs to expand the code testing to include all the systems that affect these necessary safety standards

  14. Performance Analysis of a Decoding Algorithm for Algebraic Geometry Codes

    DEFF Research Database (Denmark)

    Jensen, Helge Elbrønd; Nielsen, Rasmus Refslund; Høholdt, Tom

    1998-01-01

    We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is greater than or equal to [(dFR-1)/2]+1, where dFR is the Feng-Rao distance......We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is greater than or equal to [(dFR-1)/2]+1, where dFR is the Feng-Rao distance...

  15. 40 CFR 60.8 - Performance tests.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 6 2010-07-01 2010-07-01 false Performance tests. 60.8 Section 60.8... PERFORMANCE FOR NEW STATIONARY SOURCES General Provisions § 60.8 Performance tests. (a) Except as specified in... conduct performance test(s) and furnish the Administrator a written report of the results of such...

  16. Performance Tuning of x86 OpenMP Codes with MAQAO

    Science.gov (United States)

    Barthou, Denis; Charif Rubial, Andres; Jalby, William; Koliai, Souad; Valensi, Cédric

    Failing to find the best optimization sequence for a given application code can lead to compiler generated codes with poor performances or inappropriate code. It is necessary to analyze performances from the assembly generated code to improve over the compilation process. This paper presents a tool for the performance analysis of multithreaded codes (OpenMP programs support at the moment). MAQAO relies on static performance evaluation to identify compiler optimizations and assess performance of loops. It exploits static binary rewriting for reading and instrumenting object files or executables. Static binary instrumentation allows the insertion of probes at instruction level. Memory accesses can be captured to help tune the code, but such traces require to be compressed. MAQAO can analyze the results and provide hints for tuning the code. We show on some examples how this can help users improve their OpenMP applications.

  17. The CMSSW benchmarking suite: Using HEP code to measure CPU performance

    International Nuclear Information System (INIS)

    Benelli, G

    2010-01-01

    The demanding computing needs of the CMS experiment require thoughtful planning and management of its computing infrastructure. A key factor in this process is the use of realistic benchmarks when assessing the computing power of the different architectures available. In recent years a discrepancy has been observed between the CPU performance estimates given by the reference benchmark for HEP computing (SPECint) and actual performances of HEP code. Making use of the CPU performance tools from the CMSSW performance suite, comparative CPU performance studies have been carried out on several architectures. A benchmarking suite has been developed and integrated in the CMSSW framework, to allow computing centers and interested third parties to benchmark architectures directly with CMSSW. The CMSSW benchmarking suite can be used out of the box, to test and compare several machines in terms of CPU performance and report with the wanted level of detail the different benchmarking scores (e.g. by processing step) and results. In this talk we describe briefly the CMSSW software performance suite, and in detail the CMSSW benchmarking suite client/server design, the performance data analysis and the available CMSSW benchmark scores. The experience in the use of HEP code for benchmarking will be discussed and CMSSW benchmark results presented.

  18. Calculations of Fission Gas Release During Ramp Tests Using Copernic Code

    Energy Technology Data Exchange (ETDEWEB)

    Tong, Liu [Nuclear Fuel R and D Center, China Nuclear Power Technology Research Institute (CNPRI) (China)

    2013-03-15

    The report performed under IAEA research contract No.15951 describes the results of fuel performance evaluation of LWR fuel rods operated at ramp conditions using the COPERNIC code developed by AREVA. The experimental data from the Third Riso Fission Gas Project and the Studsvik SUPER-RAMP Project presented in the IFPE database of the OECD/NEA has been utilized for assessing the code itself during simulation of fission gas release (FGR). Standard code models for LWR fuel were used in simulations with parameters set properly in accordance with relevant test reports. With the help of data adjustment, the input power histories are restructured to fit the real ones, so as to ensure the validity of FGR prediction. The results obtained by COPERNIC show that different models lead to diverse predictions and discrepancies. By comparison, the COPERNIC V2.2 model (95% Upper bound) is selected as the standard FGR model in this report and the FGR phenomenon is properly simulated by the code. To interpret the large discrepancies of some certain PK rods, the burst effect of FGR which is taken into consideration in COPERNIC is described and the influence of the input power histories is extrapolated. In addition, the real-time tracking capability of COPERNIC is tested against experimental data. In the process of investigation, two main dominant factors influencing the measured gas release rate are described and different mechanisms are analyzed. With the limited predicting capacity, accurate predictions cannot be carried out on abrupt changes of FGR during ramp tests by COPERNIC and improvements may be necessary to some relevant models. (author)

  19. Lysimeter data as input to performance assessment source term codes

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Sullivan, T.

    1992-01-01

    The Field Lysimeter Investigation: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II c prefilters employed in the cleanup of the Three Mile Island (TMI) Nuclear Power Station are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. In this paper, radionuclide releases from waste forms in the first seven years of sampling are presented and discussed. Application of lysimeter data to be used in performance assessment source term models is presented. Initial results from use of data in two models are discussed

  20. Investigation of flashing-induced instabilities at Circus test facility with the code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Schafer, F.; Manera, A. [Forschungzentrum Rossendorf e.V., Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)]. E-mail: F.Schaefer@fz-rossendorf.de; A.Manera@fz-rossendorf.de

    2006-07-01

    The test facility CIRCUS (CIRculation Under Start-up) was built to study the start-up phase of a natural-circulation BWR. During the start-up,so-called flashing-induced instabilities can arise. These instabilities are induced by flashing (i.e., steam production in adiabatic conditions) of the coolant in the long riser section, which is placed above the core to enhance the flow rate. The flashing that occurs in the riser causes an imbalance between driving force and pressure losses in the natural-circulation loop, giving rise to flow oscillations. Within the European-Union 5th Framework Programme, a project, NACUSP (Natural circulation and stability performance of BWRs), has been started in December 2000, having as one of its main aims the understanding of the physics of the phenomena involved during the start-up phase of natural-circulation-cooled BWRs, providing a large experimental database and validating state-of-the-art thermo-hydraulic codes in the low-pressure, low-power operational region of these reactors. One part of this project deals with the modelling of selected CIRCUS tests using the thermo-hydraulic code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients). This paper gives an overview about experiments and simulations. The code ATHLET is used to investigate the dynamic behaviour of the CIRCUS test facility and the results of the calculations are compared with the experimental data. (author)

  1. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Tso, C.F. [Arup (United Kingdom); Hueggenberg, R. [Gesellschaft fuer Nuklear-Behaelter mbH (Germany)

    2004-07-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work.

  2. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    International Nuclear Information System (INIS)

    Tso, C.F.; Hueggenberg, R.

    2004-01-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work

  3. Implementation and testing of the CFDS-FLOW3D code

    International Nuclear Information System (INIS)

    Smith, B.L.

    1994-03-01

    FLOW3D is a multi-purpose, transient fluid dynamics and heat transfer code developed by Computational Fluid Dynamics Services (CFDS), a branch of AEA Technology, based at Harwell. The code is supplied with a SUN-based operating environment consisting of an interactive grid generator SOPHIA and a post-processor JASPER for graphical display of results. Both SOPHIA and JASPER are extensions of the support software originally written for the ASTEC code, also promoted by CFDS. The latest release of FLOW3D contains well-tested turbulence and combustion models and, in a less-developed form, a multi-phase modelling potential. This document describes briefly the modelling capabilities of FLOW3D (Release 3.2) and outlines implementation procedures for the VAX, CRAY and CONVEX computer systems. Additional remarks are made concerning the in-house support programs which have been specially written in order to adapt existing ASTEC input data for use with FLOW3D; these programs operate within a VAX-VMS environment. Three sample calculations have been performed and results compared with those obtained previously using the ASTEC code, and checked against other available data, where appropriate. (author) 35 figs., 3 tabs., 42 refs

  4. Implementation of computer codes for performance assessment of the Republic repository of LLW/ILW Mochovce

    International Nuclear Information System (INIS)

    Hanusik, V.; Kopcani, I.; Gedeon, M.

    2000-01-01

    This paper describes selection and adaptation of computer codes required to assess the effects of radionuclide release from Mochovce Radioactive Waste Disposal Facility. The paper also demonstrates how these codes can be integrated into performance assessment methodology. The considered codes include DUST-MS for source term release, MODFLOW for ground-water flow and BS for transport through biosphere and dose assessment. (author)

  5. Application of startup/core management code system to YGN 3 startup testing

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Hah, Yung Joon; Doo, Jin Yong; Kim, Dae Kyum

    1995-01-01

    YGN 3 is the first nuclear power plant in Korea to use the fixed incore detector system for startup testing and core management. The startup/core management code system was developed from existing ABB-C-E codes and applied for YGN 3 startup testing, especially for physics and CPC(Core Protection Calculator)/COLSS (Core Operating Limit Supervisory System) related testing. The startup/core management code system consists of startup codes which include the CEBASE, CECOR, CEFAST and CEDOPS, and startup data reduction codes which include FLOWRATE, COREPERF, CALMET, and VARTAV. These codes were implemented on an HP/Apollo model 9000 series 400 workstation at the YGN 3 site and successfully applied to startup testing and core management. The startup codes made a great contribution in upgrading the reliability of test results and reducing the test period by taking and analyzing core data automatically. The data reduction code saved the manpower and time for test data reduction and decreased the chance for error in the analysis. It is expected that this code system will make similar contributions for reducing the startup testing duration of YGN 4 and UCN3,4

  6. Assessment of predictive capability of REFLA/TRAC code for large break LOCA transient in PWR using LOFT L2-5 test data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

    1994-03-01

    The REFLA/TRAC code is a best estimate code developed at Japan Atomic Energy Research Institute (JAERI) to provide advanced predictions of thermal hydraulic transient in light water reactors (LWRs). The REFLA/TRAC code uses the TRAC-PF1/MOD1 code as the framework of the code. The REFLA/TRAC code is expected to be used for the calibration of licensing codes, accident analysis, accident simulation of LWRs, and design of advanced LWRs. Several models have been implemented to the TRAC-PF1/MOD1 code at JAERI including reflood model, condensation model, interfacial and wall friction models, etc. These models have been verified using data from various separate effect tests. This report describes an assessment result of the REFLA/TRAC code, which was performed to assess the predictive capability for integral system behavior under large break loss of coolant accident (LBLOCA) using data from the LOFT L2-5 test. The assessment calculation confirmed that the REFLA/TRAC code can predict break mass flow rate, emergency core cooling water bypass and clad temperature excellently in the LOFT L2-5 test. The CPU time of the REFLA/TRAC code was about 1/3 of the TRAC-PF1/MOD1 code. The REFLA/TRAC code can perform stable and fast simulation of thermal hydraulic behavior in PWR LBLOCA with enough accuracy for practical use. (author)

  7. Stereotype Threat, Test Anxiety, and Mathematics Performance

    Science.gov (United States)

    Tempel, Tobias; Neumann, Roland

    2014-01-01

    We investigated the combined effects of stereotype threat and trait test anxiety on mathematics test performance. Stereotype threat and test anxiety interacted with each other in affecting performance. Trait test anxiety predicted performance only in a diagnostic condition that prevented stereotype threat by stereotype denial. A state measure of…

  8. Performance Analysis of Faulty Gallager-B Decoding of QC-LDPC Codes with Applications

    Directory of Open Access Journals (Sweden)

    O. Al Rasheed

    2014-06-01

    Full Text Available In this paper we evaluate the performance of Gallager-B algorithm, used for decoding low-density parity-check (LDPC codes, under unreliable message computation. Our analysis is restricted to LDPC codes constructed from circular matrices (QC-LDPC codes. Using Monte Carlo simulation we investigate the effects of different code parameters on coding system performance, under a binary symmetric communication channel and independent transient faults model. One possible application of the presented analysis in designing memory architecture with unreliable components is considered.

  9. Performance Theories for Sentence Coding: Some Quantitative Models

    Science.gov (United States)

    Aaronson, Doris; And Others

    1977-01-01

    This study deals with the patterns of word-by-word reading times over a sentence when the subject must code the linguistic information sufficiently for immediate verbatim recall. A class of quantitative models is considered that would account for reading times at phrase breaks. (Author/RM)

  10. Performance Analysis of DPSK Signals with Selection Combining and Convolutional Coding in Fading Channel

    National Research Council Canada - National Science Library

    Ong, Choon

    1998-01-01

    The performance analysis of a differential phase shift keyed (DPSK) communications system, operating in a Rayleigh fading environment, employing convolutional coding and diversity processing is presented...

  11. High Performance Electrical Modeling and Simulation Verification Test Suite - Tier I; TOPICAL

    International Nuclear Information System (INIS)

    SCHELLS, REGINA L.; BOGDAN, CAROLYN W.; WIX, STEVEN D.

    2001-01-01

    This document describes the High Performance Electrical Modeling and Simulation (HPEMS) Global Verification Test Suite (VERTS). The VERTS is a regression test suite used for verification of the electrical circuit simulation codes currently being developed by the HPEMS code development team. This document contains descriptions of the Tier I test cases

  12. Performance of automated and manual coding systems for occupational data: a case study of historical records.

    Science.gov (United States)

    Patel, Mehul D; Rose, Kathryn M; Owens, Cindy R; Bang, Heejung; Kaufman, Jay S

    2012-03-01

    Occupational data are a common source of workplace exposure and socioeconomic information in epidemiologic research. We compared the performance of two occupation coding methods, an automated software and a manual coder, using occupation and industry titles from U.S. historical records. We collected parental occupational data from 1920-40s birth certificates, Census records, and city directories on 3,135 deceased individuals in the Atherosclerosis Risk in Communities (ARIC) study. Unique occupation-industry narratives were assigned codes by a manual coder and the Standardized Occupation and Industry Coding software program. We calculated agreement between coding methods of classification into major Census occupational groups. Automated coding software assigned codes to 71% of occupations and 76% of industries. Of this subset coded by software, 73% of occupation codes and 69% of industry codes matched between automated and manual coding. For major occupational groups, agreement improved to 89% (kappa = 0.86). Automated occupational coding is a cost-efficient alternative to manual coding. However, some manual coding is required to code incomplete information. We found substantial variability between coders in the assignment of occupations although not as large for major groups.

  13. Test and intercomparisons of data fitting with general least squares code GMA versus Bayesian code GLUCS

    International Nuclear Information System (INIS)

    Pronyaev, V.G.

    2003-01-01

    Data fitting with GMA and GLUCS gives consistent results. Difference in the evaluated central values obtained with different formalisms can be related to the general accuracy with which fits could be done in different formalisms. It has stochastic nature and should be accounted in the final results of the data evaluation as small SERC uncertainty. Some shift in central values of data evaluated with GLUCS and GMA relative the central values evaluated with the R-matrix model code RAC is observed for cases of fitting strongly varying data and is related to the PPP. The procedure of evaluation, free from PPP, should be elaborated. (author)

  14. Analysis of L test series of ACE (Advanced Containment Experiments) project with modified corcon UW code

    International Nuclear Information System (INIS)

    Laguna Velasco, H.

    1994-01-01

    A series of experimental tests (so call L, Large scale) have been performance under sponsored of many research institutions around the world and management by Electric Power Research Institute at U.S.A. The goal of these tests is to analyze the phenomena of core-concrete interaction at the same conditions as severe accident in light water nuclear reactor. Results of these tests provides experimental data about thermohydraulic phenomenon and aerosol and fission products release. With these results, improves many codes that already have been developed to simulate core-concrete interaction during severe accident ; in case of CORCON.UW code is a improved version developed in University of Wisconsin at CORCON MOD 2. Scope of this work is shown results obtained from CORCON.UW improved. The improves consist of add data about BaSiO 3 , Ba 2 SiO 4 , BaZrO 3 , SrSiO 4 and SrZrO 3 , append Kutateladze's heat transfer correlation, and finally make more efficient the resolution of energy equations system through use a better algorithm. The results obtained by this improved code to the downward power and H 2 , H 2 O, CO and CO 2 release are agree with experimental results, and also it saved 40% of C.P.U. consumption during execution, due improve of energy equation system. Conclusions are, the increase of thermodynamics data in CORCON.UW produce a well results comparative with experimental results and update heat transfer correlations and algorithm brings a versatile code and reliable results. (Author)

  15. Performance evaluations of advanced massively parallel platforms based on gyrokinetic toroidal five-dimensional Eulerian code GT5D

    International Nuclear Information System (INIS)

    Idomura, Yasuhiro; Jolliet, Sebastien

    2010-01-01

    A gyrokinetic toroidal five dimensional Eulerian code GT5D is ported on six advanced massively parallel platforms and comprehensive benchmark tests are performed. A parallelisation technique based on physical properties of the gyrokinetic equation is presented. By extending the parallelisation technique with a hybrid parallel model, the scalability of the code is improved on platforms with multi-core processors. In the benchmark tests, a good salability is confirmed up to several thousands cores on every platforms, and the maximum sustained performance of ∼18.6 Tflops is achieved using 16384 cores of BX900. (author)

  16. Enhanced verification test suite for physics simulation codes

    Energy Technology Data Exchange (ETDEWEB)

    Kamm, James R.; Brock, Jerry S.; Brandon, Scott T.; Cotrell, David L.; Johnson, Bryan; Knupp, Patrick; Rider, William J.; Trucano, Timothy G.; Weirs, V. Gregory

    2008-09-01

    This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations.

  17. Calculation of Sodium Fire Test-I (Run-E6) using sodium combustion analysis code ASSCOPS version 2.0

    Energy Technology Data Exchange (ETDEWEB)

    Nakagiri, Toshio; Ohno, Shuji; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-11-01

    The calculation of Sodium Fire Test-I (Run-E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because calculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results. (author)

  18. Performance enhancement of successive interference cancellation scheme based on spectral amplitude coding for optical code-division multiple-access systems using Hadamard codes

    Science.gov (United States)

    Eltaif, Tawfig; Shalaby, Hossam M. H.; Shaari, Sahbudin; Hamarsheh, Mohammad M. N.

    2009-04-01

    A successive interference cancellation scheme is applied to optical code-division multiple-access (OCDMA) systems with spectral amplitude coding (SAC). A detailed analysis of this system, with Hadamard codes used as signature sequences, is presented. The system can easily remove the effect of the strongest signal at each stage of the cancellation process. In addition, simulation of the prose system is performed in order to validate the theoretical results. The system shows a small bit error rate at a large number of active users compared to the SAC OCDMA system. Our results reveal that the proposed system is efficient in eliminating the effect of the multiple-user interference and in the enhancement of the overall performance.

  19. CENER/NREL Collaboration in Testing Facility and Code Development: Cooperative Research and Development Final Report, CRADA Number CRD-06-207

    Energy Technology Data Exchange (ETDEWEB)

    Moriarty, P.

    2014-11-01

    Under the funds-in CRADA agreement, NREL and CENER will collaborate in the areas of blade and drivetrain testing facility development and code development. The project shall include NREL assisting in the review and instruction necessary to assist in commissioning the new CENER blade test and drivetrain test facilities. In addition, training will be provided by allowing CENER testing staff to observe testing and operating procedures at the NREL blade test and drivetrain test facilities. CENER and NREL will exchange blade and drivetrain facility and equipment design and performance information. The project shall also include exchanging expertise in code development and data to validate numerous computational codes.

  20. Operator performance in non-destructive testing: A study of operator performance in a performance test

    Energy Technology Data Exchange (ETDEWEB)

    Enkvist, J.; Edland, A.; Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology

    2000-05-15

    In the process industries there is a need of inspecting the integrity of critical components without disrupting the process. Such in-service inspections are typically performed with non-destructive testing (NDT). In NDT the task of the operator is to (based on diagnostic information) decide if the component can remain in service or not. The present study looks at the performance in NDT. The aim is to improve performance, in the long run, by exploring the operators' decision strategies and other underlying factors and to this way find out what makes some operators more successful than others. Sixteen operators performed manual ultrasonic inspections of four test pieces with the aim to detect (implanted) cracks. In addition to these performance demonstration tests (PDT), the operators performed independent ability tests and filled out questionnaires. The results show that operators who trust their gut feeling more than the procedure (when the two come to different results) and that at the same time have a positive attitude towards the procedure have a higher PDT performance. These results indicate the need for operators to be motivated and confident when performing NDT. It was also found that the operators who performed better rated more decision criteria higher in the detection phase than the operators who performed worse. For characterizing it was the other way around. Also, the operators who performed better used more time, both detecting and characterizing, than the operators who performed worse.

  1. Operator performance in non-destructive testing: A study of operator performance in a performance test

    International Nuclear Information System (INIS)

    Enkvist, J.; Edland, A.; Svenson, Ola

    2000-05-01

    In the process industries there is a need of inspecting the integrity of critical components without disrupting the process. Such in-service inspections are typically performed with non-destructive testing (NDT). In NDT the task of the operator is to (based on diagnostic information) decide if the component can remain in service or not. The present study looks at the performance in NDT. The aim is to improve performance, in the long run, by exploring the operators' decision strategies and other underlying factors and to this way find out what makes some operators more successful than others. Sixteen operators performed manual ultrasonic inspections of four test pieces with the aim to detect (implanted) cracks. In addition to these performance demonstration tests (PDT), the operators performed independent ability tests and filled out questionnaires. The results show that operators who trust their gut feeling more than the procedure (when the two come to different results) and that at the same time have a positive attitude towards the procedure have a higher PDT performance. These results indicate the need for operators to be motivated and confident when performing NDT. It was also found that the operators who performed better rated more decision criteria higher in the detection phase than the operators who performed worse. For characterizing it was the other way around. Also, the operators who performed better used more time, both detecting and characterizing, than the operators who performed worse

  2. PERCON: A flexible computer code for detailed thermal performance studies

    International Nuclear Information System (INIS)

    Boardman, F.B.; Collier, W.D.

    1975-07-01

    PERCON is a computer code which evaluates temperatures in three dimensions for a block containing heat sources and having coolant flow in one dimension. The solution is obtained at successive planes perpendicular to the coolant flow and the progression from one plane to the next occurs by the heat to the coolant determining convective boundary conditions at the next plane after due allowance being made for any lateral mixing or mass transfer between coolants. It is also possible to calculate the diametral change along a radius as a function of irradiation shrinkage and thermal expansion. This is used in a 'through life' calculation which evalates interaction pressure in tubular fuel elements. Physical property data used by the code may be specified as functions of temperature. The coolant flow may be specified, or alternatively derived by the program to satisfy either a specified overall pressure drop or mixed mean temperature rise. The pressure drop through each coolant is calculated and the flow modified, followed by a repeat of the temperature calculation, until the pressure imbalance between chosen flow channels at chosen axial positions is less than the specified convergence limit. A detailed description of the facilities in the code is given and some cases which have been studied are discussed. (U.K.)

  3. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S.

    2002-05-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  4. Ecotoxicological testing of performance fluids

    International Nuclear Information System (INIS)

    Kallqvist, T.

    1990-05-01

    The report deals with a project comprising the testing of drilling fluids concerning ecotoxicology, biological degradation, and toxicity. Two types of drilling fluids were tested for toxic effects on marine algae and biological degradability. A fluid based on mineral oil was readily degradable (98% DOC removal in 28 days) while an ether based oil degraded more slowly (56% DOC removal in 28 days). The toxicity of both fluids was tested after emulsification of the oils in water and separating the oil and water phase after equilibration. The EC 50 values obtained with this approach were 8.15 g/l for the oil based fluid and 116 g/l for the ether fluid. 9 figs., 8 tabs

  5. Gest-sip1 experiments and post-test calculations with the relap5 code

    International Nuclear Information System (INIS)

    Achilli, A.; Cattadori, G.; Ferri, R.; Gandolfi, S.; Bianchi, F.; Meloni, P.

    2001-01-01

    The SIP-1 apparatus (Sistema di Iniezione Passiva) was conceived, designed, numerically simulated and tested by the SIET company as an innovative depressurization and make-up device for the New Generation LWRs. In particular it is suitable to cope with those accidents where pressure in the circuit must be dumped to allow low pressure injection systems to intervene. The main peculiarity of SIP-1 is the capability of de-pressurizing a system by cold water injection, rather than by discharging mass to the outlet, as in the common depressurization systems. ENEA sponsored all the research activity, starting from the SIP-1 design, its numerical simulation with the Relap5 code, the realisation of an experimental facility up to the test execution and post-test calculations. An experimental campaign on the GEST-SIP1 facility was performed in July 2000. The facility is mainly constituted by a U-tube Steam Generator which a proper model of SIP-1 apparatus is connected to. A series of Small Break LOCAs was simulated by varying the break size and different steady conditions were investigated to verify the stability of SIP-1, the lack of unexpected interventions and the actuation modalities. This paper deals with the description of the GEST-SIP1 experimental facility, the SIP-1 operating principles, the most meaningful results of the tests and the capability of the Relap5 code in reproducing phenomena and events. (author)

  6. Calculations to an IAHR-benchmark test using the CFD-code CFX-4

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E

    1998-10-01

    The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)

  7. Performance analysis of wavelength/spatial coding system with fixed in-phase code matrices in OCDMA network

    Science.gov (United States)

    Tsai, Cheng-Mu; Liang, Tsair-Chun

    2011-12-01

    This paper proposes a wavelength/spatial (W/S) coding system with fixed in-phase code (FIPC) matrix in the optical code-division multiple-access (OCDMA) network. A scheme is presented to form the FIPC matrix which is applied to construct the W/S OCDMA network. The encoder/decoder in the W/S OCDMA network is fully able to eliminate the multiple-access-interference (MAI) at the balanced photo-detectors (PD), according to fixed in-phase cross correlation. The phase-induced intensity noise (PIIN) related to the power square is markedly suppressed in the receiver by spreading the received power into each PD while the net signal power is kept the same. Simulation results show that the W/S OCDMA network based on the FIPC matrices cannot only completely remove the MAI but effectively suppress the PIIN to upgrade the network performance.

  8. Counter-part Test and Code Analysis of the Integral Test Loop, SNUF

    International Nuclear Information System (INIS)

    Park, Goon Cherl; Bae, B. U.; Lee, K. H.; Cho, Y. J.

    2007-02-01

    The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in pressurized water reactor, APR1400, were investigated. The reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment of SNUF, the energy scaling methodology was suggested as scaling the coolant mass inventory and thermal power for the reduced-pressure condition. From the MARS code analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. The experimental results was utilized to validate the calculation capability of MARS

  9. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  10. Knowledge and Performance about Nursing Ethic Codes from Nurses' and Patients' Perspective in Tabriz Teaching Hospitals, Iran

    Directory of Open Access Journals (Sweden)

    Sara Moghaddam

    2013-08-01

    Full Text Available Introduction: Nursing profession requires knowledge of ethics to guide performance. The nature of this profession necessitates ethical care more than routine care. Today, worldwide definition of professional ethic code has been done based on human and ethical issues in the communication between nurse and patient. To improve all dimensions of nursing, we need to respect ethic codes. The aim of this study is to assess knowledge and performance about nursing ethic codes from nurses' and patients' perspective.Methods: A cross-sectional comparative study Conducted upon 345 nurses and 500 inpatients in six teaching hospitals of Tabriz, 2012. To investigate nurses' knowledge and performance, data were collected by using structured questionnaires. Statistical analysis was done using descriptive and analytic statistics, independent t-test and ANOVA and Pearson correlation coefficient, in SPSS13.Results: Most of the nurses were female, married, educated at BS degree and 86.4% of them were aware of Ethic codes also 91.9% of nurses and 41.8% of patients represented nurses respect ethic codes. Nurses' and patients' perspective about ethic codes differed significantly. Significant relationship was found between nurses' knowledge of ethic codes and job satisfaction and complaint of ethical performance. Conclusion: According to the results, consideration to teaching ethic codes in nursing curriculum for student and continuous education for staff is proposed, on the other hand recognizing failures of the health system, optimizing nursing care, attempt to inform patients about Nursing ethic codes, promote patient rights and achieve patient satisfaction can minimize the differences between the two perspectives.

  11. The Barrier code for predicting long-term concrete performance

    International Nuclear Information System (INIS)

    Shuman, R.; Rogers, V.C.; Shaw, R.A.

    1989-01-01

    There are numerous features incorporated into a LLW disposal facility that deal directly with critical safety objectives required by the NRC in 10 CFR 61. Engineered barriers or structures incorporating concrete are commonly being considered for waste disposal facilities. The Barrier computer code calculates the long-term degradation of concrete structures in LLW disposal facilities. It couples this degradation with water infiltration into the facility, nuclide leaching from the waste, contaminated water release from the facility, and associated doses to members of the critical population group. The concrete degradation methodology of Barrier is described

  12. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  13. A comparison of thermal algorithms of fuel rod performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

  14. Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

    Energy Technology Data Exchange (ETDEWEB)

    Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2016-05-01

    In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. In particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.

  15. ENDF/B Pre-Processing Codes: Implementing and testing on a Personal Computer

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1987-05-01

    This document describes the contents of the diskettes containing the ENDF/B Pre-Processing codes by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a series of 7 diskettes. (author)

  16. Analysis of PHEBUS FPT1 test with IMPACT/SAMPSON code

    International Nuclear Information System (INIS)

    Terada, Masafumi; Ikeda, Takashi; Naitoh, Masanori

    2003-01-01

    IMPACT is a simulation software developed at the Nuclear Power Engineering Corporation, which includes the severe accident analysis code, SAMPSON. SAMPSON consists of twelve modules and is capable of simulating hypothesized severe accidents in LWR. Phebus-FPT1 test, which was selected as the International Standard Problem-46, was analyzed with SAMPSON for the verification of the code. The Phebus-FPT1 test was an integral in-pile experiment for studying mainly degradation of fuel bundle and subsequent FP behavior under a LWR severe accident condition, using irradiated fuel as a source of real FP. The following analyses of the Phebus-FPT1 test, which are also the subjects of the ISP-46, were performed: (1) In-core thermal hydraulics, core degradation and FP release from the fuel, (2) FP gas and aerosol transport in the primary circuit, (3) Thermal hydraulics and FP aerosol physics in the containment and (4) Iodine chemistry in the containment. The analysis results of the thermal hydraulics and core degradation showed good agreement with experimental data, except shroud temperatures which were higher than the experiment. The difference may be due to insufficient modeling of the gap closure in the shroud. FP release from fuel, FP transport rate in the primary circuit, FP aerosol physics and iodine chemistry in the containment were also well predicted. Through the analyses, the modules of SAMPSON used were proved to be capable for evaluating thermal hydraulics and FP behaviors under LWR severe accident conditions

  17. Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

    International Nuclear Information System (INIS)

    Risken, Tobias; Koch, Marco K.

    2011-01-01

    The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

  18. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders—attention to detail and systemizing—may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings. PMID:27242491

  19. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants.

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders-attention to detail and systemizing-may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings.

  20. Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

    International Nuclear Information System (INIS)

    Austregesilo, Henrique; Bals, Christine; Trambauer, Klaus

    2007-01-01

    In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B 4 C oxidation do not affect significantly the total calculated hydrogen release rates

  1. Construction and performance analysis of variable-weight optical orthogonal codes for asynchronous OCDMA systems

    Science.gov (United States)

    Li, Chuan-qi; Yang, Meng-jie; Zhang, Xiu-rong; Chen, Mei-juan; He, Dong-dong; Fan, Qing-bin

    2014-07-01

    A construction scheme of variable-weight optical orthogonal codes (VW-OOCs) for asynchronous optical code division multiple access (OCDMA) system is proposed. According to the actual situation, the code family can be obtained by programming in Matlab with the given code weight and corresponding capacity. The formula of bit error rate (BER) is derived by taking account of the effects of shot noise, avalanche photodiode (APD) bulk, thermal noise and surface leakage currents. The OCDMA system with the VW-OOCs is designed and improved. The study shows that the VW-OOCs have excellent performance of BER. Despite of coming from the same code family or not, the codes with larger weight have lower BER compared with the other codes in the same conditions. By taking simulation, the conclusion is consistent with the analysis of BER in theory. And the ideal eye diagrams are obtained by the optical hard limiter.

  2. Online test application development using framework CodeIgniter

    Science.gov (United States)

    Wibawa, S. C.; Wahyuningsih, Y.; Sulistyowati, R.; Abidin, R.; Lestari, Y.; Noviyanti; Maulana, D. A.

    2018-01-01

    The purpose of this study is developing application an online test for vocational students and to know the user acceptance testing on the application. The method used in this research is the Research and Development (R & D) only up to the pilot phase of the product. The stage of the procedure of the research namely: (1) Analyze the exam using paper compared to using web-based application test online. (2) Design the media in accordance with the design of the author. (3) To test the product by including a questionnaire instrument against the application that has been done. Researchers carried out tests on class X on the computer and network engineering Vocational High School (SMK) Darul Ma’wa Plumpang. It can be concluded that: (1) application online test was created gets the value of the validator with the percentage of lowest value and the highest value for the validation of products: 25% and 100%. With a total number of 14 questions, after validation of the products obtained from the three aspects of the assessment scale from 81.25 to 100 obtained from 2 different validators with the meaning of an application that has been developed and very suitable for use in school. (2) Based on User Acceptance Testing (UAT), applications can be very well received by the students and recommend to replay the final semester and others. With the successful acquisition of a category which means it’s ready and qualified.

  3. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  4. Development of Pflotran Code for Waste Isolation Pilot Plant Performance Assessment

    Science.gov (United States)

    Zeitler, T.; Day, B. A.; Frederick, J.; Hammond, G. E.; Kim, S.; Sarathi, R.; Stein, E.

    2017-12-01

    The Waste Isolation Pilot Plant (WIPP) has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. Containment of TRU waste at the WIPP is regulated by the U.S. Environmental Protection Agency (EPA). The DOE demonstrates compliance with the containment requirements by means of performance assessment (PA) calculations. WIPP PA calculations estimate the probability and consequence of potential radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. The long-term performance of the repository is assessed using a suite of sophisticated computational codes. There is a current effort to enhance WIPP PA capabilities through the further development of the PFLOTRAN software, a state-of-the-art massively parallel subsurface flow and reactive transport code. Benchmark testing of the individual WIPP-specific process models implemented in PFLOTRAN (e.g., gas generation, chemistry, creep closure, actinide transport, and waste form) has been performed, including results comparisons for PFLOTRAN and existing WIPP PA codes. Additionally, enhancements to the subsurface hydrologic flow mode have been made. Repository-scale testing has also been performed for the modified PFLTORAN code and detailed results will be presented. Ultimately, improvements to the current computational environment will result in greater detail and flexibility in the repository model due to a move from a two-dimensional calculation grid to a three-dimensional representation. The result of the effort will be a state-of-the-art subsurface flow and transport capability that will serve WIPP PA into the future for use in compliance recertification applications (CRAs) submitted to the EPA. Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of

  5. New Technique for Improving Performance of LDPC Codes in the Presence of Trapping Sets

    Directory of Open Access Journals (Sweden)

    Mohamed Adnan Landolsi

    2008-06-01

    Full Text Available Trapping sets are considered the primary factor for degrading the performance of low-density parity-check (LDPC codes in the error-floor region. The effect of trapping sets on the performance of an LDPC code becomes worse as the code size decreases. One approach to tackle this problem is to minimize trapping sets during LDPC code design. However, while trapping sets can be reduced, their complete elimination is infeasible due to the presence of cycles in the underlying LDPC code bipartite graph. In this work, we introduce a new technique based on trapping sets neutralization to minimize the negative effect of trapping sets under belief propagation (BP decoding. Simulation results for random, progressive edge growth (PEG and MacKay LDPC codes demonstrate the effectiveness of the proposed technique. The hardware cost of the proposed technique is also shown to be minimal.

  6. The Uncertainty Test for the MAAP Computer Code

    International Nuclear Information System (INIS)

    Park, S. H.; Song, Y. M.; Park, S. Y.; Ahn, K. I.; Kim, K. R.; Lee, Y. J.

    2008-01-01

    After the Three Mile Island Unit 2 (TMI-2) and Chernobyl accidents, safety issues for a severe accident are treated in various aspects. Major issues in our research part include a level 2 PSA. The difficulty in expanding the level 2 PSA as a risk information activity is the uncertainty. In former days, it attached a weight to improve the quality in a internal accident PSA, but the effort is insufficient for decrease the phenomenon uncertainty in the level 2 PSA. In our country, the uncertainty degree is high in the case of a level 2 PSA model, and it is necessary to secure a model to decrease the uncertainty. We have not yet experienced the uncertainty assessment technology, the assessment system itself depends on advanced nations. In advanced nations, the severe accident simulator is implemented in the hardware level. But in our case, basic function in a software level can be implemented. In these circumstance at home and abroad, similar instances are surveyed such as UQM and MELCOR. Referred to these instances, SAUNA (Severe Accident UNcertainty Analysis) system is being developed in our project to assess and decrease the uncertainty in a level 2 PSA. It selects the MAAP code to analyze the uncertainty in a severe accident

  7. A Test of Two Alternative Cognitive Processing Models: Learning Styles and Dual Coding

    Science.gov (United States)

    Cuevas, Joshua; Dawson, Bryan L.

    2018-01-01

    This study tested two cognitive models, learning styles and dual coding, which make contradictory predictions about how learners process and retain visual and auditory information. Learning styles-based instructional practices are common in educational environments despite a questionable research base, while the use of dual coding is less…

  8. A Correlational Study: Code of Ethics in Testing and EFL Instructors' Professional Behavior

    Science.gov (United States)

    Ashraf, Hamid; Kafi, Zahra; Saeedan, Azaam

    2018-01-01

    The present study has aimed at delving the code of ethics in testing in English language institutions to see how far adhering to these ethical codes will result in EFL teachers' professional behavior. Therefore, 300 EFL instructors teaching at English language schools in Khorasan Razavi Province, Zabansara Language School, as well as Khorasan…

  9. High-Performance Java Codes for Computational Fluid Dynamics

    Science.gov (United States)

    Riley, Christopher; Chatterjee, Siddhartha; Biswas, Rupak; Biegel, Bryan (Technical Monitor)

    2001-01-01

    The computational science community is reluctant to write large-scale computationally -intensive applications in Java due to concerns over Java's poor performance, despite the claimed software engineering advantages of its object-oriented features. Naive Java implementations of numerical algorithms can perform poorly compared to corresponding Fortran or C implementations. To achieve high performance, Java applications must be designed with good performance as a primary goal. This paper presents the object-oriented design and implementation of two real-world applications from the field of Computational Fluid Dynamics (CFD): a finite-volume fluid flow solver (LAURA, from NASA Langley Research Center), and an unstructured mesh adaptation algorithm (2D_TAG, from NASA Ames Research Center). This work builds on our previous experience with the design of high-performance numerical libraries in Java. We examine the performance of the applications using the currently available Java infrastructure and show that the Java version of the flow solver LAURA performs almost within a factor of 2 of the original procedural version. Our Java version of the mesh adaptation algorithm 2D_TAG performs within a factor of 1.5 of its original procedural version on certain platforms. Our results demonstrate that object-oriented software design principles are not necessarily inimical to high performance.

  10. NGNP Component Test Capability Design Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad; D.S. Ferguson; L.E. Guillen; C.W. McKnight; P.J. Petersen

    2009-09-01

    The Next Generation Nuclear Plant Project is conducting a trade study to select a preferred approach for establishing a capability whereby NGNP technology development testing—through large-scale, integrated tests—can be performed for critical HTGR structures, systems, and components (SSCs). The mission of this capability includes enabling the validation of interfaces, interactions, and performance for critical systems and components prior to installation in the NGNP prototype.

  11. Simulation of IST Turbomachinery Power-Neutral Tests with the ANL Plant Dynamics Code

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-13

    The validation of the Plant Dynamics Code (PDC) developed at Argonne National Laboratory (ANL) for the steady-state and transient analysis of supercritical carbon dioxide (sCO2) systems has been continued with new test data from the Naval Nuclear Laboratory (operated by Bechtel Marine Propulsion Corporation) Integrated System Test (IST). Although data from three runs were provided to ANL, only two of the data sets were analyzed and described in this report. The common feature of these tests is the power-neutral operation of the turbine-compressor shaft, where no external power through the alternator was provided during the tests. Instead, the shaft speed was allowed to change dictated by the power balance between the turbine, the compressor, and the power losses in the shaft. The new test data turned out to be important for code validation for several reasons. First, the power-neutral operation of the shaft allows validation of the shaft dynamics equations in asynchronous mode, when the shaft is disconnected from the grid. Second, the shaft speed control with the compressor recirculation (CR) valve not only allows for testing the code control logic itself, but it also serves as a good test for validation of both the compressor surge control and the turbine bypass control actions, since the effect of the CR action on the loop conditions is similar for both of these controls. Third, the varying compressor-inlet temperature change test allows validation of the transient response of the precooler, a shell-and-tube heat exchanger. The first transient simulation of the compressor-inlet temperature variation Test 64661 showed a much slower calculated response of the precooler in the calculations than the test data. Further investigation revealed an error in calculating the heat exchanger tube mass for the PDC dynamic equations that resulted in a slower change in the tube wall temperature than measured. The transient calculations for both tests were done in two steps. The

  12. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  13. Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez

    2015-07-15

    Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

  14. Performance analysis of multiple interference suppression over asynchronous/synchronous optical code-division multiple-access system based on complementary/prime/shifted coding scheme

    Science.gov (United States)

    Nieh, Ta-Chun; Yang, Chao-Chin; Huang, Jen-Fa

    2011-08-01

    A complete complementary/prime/shifted prime (CPS) code family for the optical code-division multiple-access (OCDMA) system is proposed. Based on the ability of complete complementary (CC) code, the multiple-access interference (MAI) can be suppressed and eliminated via spectral amplitude coding (SAC) OCDMA system under asynchronous/synchronous transmission. By utilizing the shifted prime (SP) code in the SAC scheme, the hardware implementation of encoder/decoder can be simplified with a reduced number of optical components, such as arrayed waveguide grating (AWG) and fiber Bragg grating (FBG). This system has a superior performance as compared to previous bipolar-bipolar coding OCDMA systems.

  15. DEXTER: A one-dimensional code for calculating thermionic performance of long converters

    Science.gov (United States)

    Sawyer, C. D.

    1971-01-01

    A versatile code is described for computing the coupled thermionic electric-thermal performance of long thermionic converters in which the temperature and voltage variations cannot be neglected. The code is capable of accounting for a variety of external electrical connection schemes, coolant flow paths and converter failures by partial shorting. Example problem solutions are included along with a user's manual.

  16. Dexter - A one-dimensional code for calculating thermionic performance of long converters.

    Science.gov (United States)

    Sawyer, C. D.

    1971-01-01

    This paper describes a versatile code for computing the coupled thermionic electric-thermal performance of long thermionic converters in which the temperature and voltage variations cannot be neglected. The code is capable of accounting for a variety of external electrical connection schemes, coolant flow paths and converter failures by partial shorting. Example problem solutions are given.

  17. Light water reactor pressure isolation valve performance testing

    International Nuclear Information System (INIS)

    Neely, H.H.; Jeanmougin, N.M.; Corugedo, J.J.

    1990-07-01

    The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (author)

  18. Integrated Performance Testing Workshop, Modules 6 - 11

    Energy Technology Data Exchange (ETDEWEB)

    Leach, Janice; Torres, Teresa M.

    2012-10-01

    These modules cover performance testing of: Interior Detection Systems; Access Controls; Exterior Detection Systems; Video Assessment Systems; SNM / Contraband Detection Systems; Access Delay Elements

  19. Numerical relativity for D dimensional axially symmetric space-times: Formalism and code tests

    International Nuclear Information System (INIS)

    Zilhao, Miguel; Herdeiro, Carlos; Witek, Helvi; Nerozzi, Andrea; Sperhake, Ulrich; Cardoso, Vitor; Gualtieri, Leonardo

    2010-01-01

    The numerical evolution of Einstein's field equations in a generic background has the potential to answer a variety of important questions in physics: from applications to the gauge-gravity duality, to modeling black hole production in TeV gravity scenarios, to analysis of the stability of exact solutions, and to tests of cosmic censorship. In order to investigate these questions, we extend numerical relativity to more general space-times than those investigated hitherto, by developing a framework to study the numerical evolution of D dimensional vacuum space-times with an SO(D-2) isometry group for D≥5, or SO(D-3) for D≥6. Performing a dimensional reduction on a (D-4) sphere, the D dimensional vacuum Einstein equations are rewritten as a 3+1 dimensional system with source terms, and presented in the Baumgarte, Shapiro, Shibata, and Nakamura formulation. This allows the use of existing 3+1 dimensional numerical codes with small adaptations. Brill-Lindquist initial data are constructed in D dimensions and a procedure to match them to our 3+1 dimensional evolution equations is given. We have implemented our framework by adapting the Lean code and perform a variety of simulations of nonspinning black hole space-times. Specifically, we present a modified moving puncture gauge, which facilitates long-term stable simulations in D=5. We further demonstrate the internal consistency of the code by studying convergence and comparing numerical versus analytic results in the case of geodesic slicing for D=5, 6.

  20. PERFORMANCE ANALYSIS OF OPTICAL CDMA SYSTEM USING VC CODE FAMILY UNDER VARIOUS OPTICAL PARAMETERS

    Directory of Open Access Journals (Sweden)

    HASSAN YOUSIF AHMED

    2012-06-01

    Full Text Available The intent of this paper is to study the performance of spectral-amplitude coding optical code-division multiple-access (OCDMA systems using Vector Combinatorial (VC code under various optical parameters. This code can be constructed by an algebraic way based on Euclidian vectors for any positive integer number. One of the important properties of this code is that the maximum cross-correlation is always one which means that multi-user interference (MUI and phase induced intensity noise are reduced. Transmitter and receiver structures based on unchirped fiber Bragg grating (FBGs using VC code and taking into account effects of the intensity, shot and thermal noise sources is demonstrated. The impact of the fiber distance effects on bit error rate (BER is reported using a commercial optical systems simulator, virtual photonic instrument, VPITM. The VC code is compared mathematically with reported codes which use similar techniques. We analyzed and characterized the fiber link, received power, BER and channel spacing. The performance and optimization of VC code in SAC-OCDMA system is reported. By comparing the theoretical and simulation results taken from VPITM, we have demonstrated that, for a high number of users, even if data rate is higher, the effective power source is adequate when the VC is used. Also it is found that as the channel spacing width goes from very narrow to wider, the BER decreases, best performance occurs at a spacing bandwidth between 0.8 and 1 nm. We have shown that the SAC system utilizing VC code significantly improves the performance compared with the reported codes.

  1. Relaxation of inservice test frequency requirement for Kori 1 ASME code pumps

    International Nuclear Information System (INIS)

    Sohn, Gap Heon; Choi, Hae Yoon; Min, Kyung Sung; Rim, Nam Jin

    1994-08-01

    The objective of this investigation is to evaluate the technical and regulational requirements to justify the relaxation of the test frequency of Kori 1 pumps through reviewing the related rules and codes and standards, technical specifications of Kori 1 and other similar plants, standard technical specifications, research results for tech. spec. improvements and site test records. It is concluded that the relaxation of test frequency to quarterly be justified based on the conformance with rules and codes and standard, quarterly test cases in similar plants and standard tech. spec., recommendations of research result and stable site test records. (Author) 16 refs., 26 figs., 13 tabs

  2. Performance of the OVERFLOW-MLP and LAURA-MLP CFD Codes on the NASA Ames 512 CPU Origin System

    Science.gov (United States)

    Taft, James R.

    2000-01-01

    The shared memory Multi-Level Parallelism (MLP) technique, developed last year at NASA Ames has been very successful in dramatically improving the performance of important NASA CFD codes. This new and very simple parallel programming technique was first inserted into the OVERFLOW production CFD code in FY 1998. The OVERFLOW-MLP code's parallel performance scaled linearly to 256 CPUs on the NASA Ames 256 CPU Origin 2000 system (steger). Overall performance exceeded 20.1 GFLOP/s, or about 4.5x the performance of a dedicated 16 CPU C90 system. All of this was achieved without any major modification to the original vector based code. The OVERFLOW-MLP code is now in production on the inhouse Origin systems as well as being used offsite at commercial aerospace companies. Partially as a result of this work, NASA Ames has purchased a new 512 CPU Origin 2000 system to further test the limits of parallel performance for NASA codes of interest. This paper presents the performance obtained from the latest optimization efforts on this machine for the LAURA-MLP and OVERFLOW-MLP codes. The Langley Aerothermodynamics Upwind Relaxation Algorithm (LAURA) code is a key simulation tool in the development of the next generation shuttle, interplanetary reentry vehicles, and nearly all "X" plane development. This code sustains about 4-5 GFLOP/s on a dedicated 16 CPU C90. At this rate, expected workloads would require over 100 C90 CPU years of computing over the next few calendar years. It is not feasible to expect that this would be affordable or available to the user community. Dramatic performance gains on cheaper systems are needed. This code is expected to be perhaps the largest consumer of NASA Ames compute cycles per run in the coming year.The OVERFLOW CFD code is extensively used in the government and commercial aerospace communities to evaluate new aircraft designs. It is one of the largest consumers of NASA supercomputing cycles and large simulations of highly resolved full

  3. Bearing performance degradation assessment based on time-frequency code features and SOM network

    International Nuclear Information System (INIS)

    Zhang, Yan; Tang, Baoping; Han, Yan; Deng, Lei

    2017-01-01

    Bearing performance degradation assessment and prognostics are extremely important in supporting maintenance decision and guaranteeing the system’s reliability. To achieve this goal, this paper proposes a novel feature extraction method for the degradation assessment and prognostics of bearings. Features of time-frequency codes (TFCs) are extracted from the time-frequency distribution using a hybrid procedure based on short-time Fourier transform (STFT) and non-negative matrix factorization (NMF) theory. An alternative way to design the health indicator is investigated by quantifying the similarity between feature vectors using a self-organizing map (SOM) network. On the basis of this idea, a new health indicator called time-frequency code quantification error (TFCQE) is proposed to assess the performance degradation of the bearing. This indicator is constructed based on the bearing real-time behavior and the SOM model that is previously trained with only the TFC vectors under the normal condition. Vibration signals collected from the bearing run-to-failure tests are used to validate the developed method. The comparison results demonstrate the superiority of the proposed TFCQE indicator over many other traditional features in terms of feature quality metrics, incipient degradation identification and achieving accurate prediction. Highlights • Time-frequency codes are extracted to reflect the signals’ characteristics. • SOM network served as a tool to quantify the similarity between feature vectors. • A new health indicator is proposed to demonstrate the whole stage of degradation development. • The method is useful for extracting the degradation features and detecting the incipient degradation. • The superiority of the proposed method is verified using experimental data. (paper)

  4. The Development of Computer Code for Safety Injection Tank (SIT) with Fluidic Device(FD) Blowdown Test

    International Nuclear Information System (INIS)

    Lee, Joo Hee; Kim, Tae Han; Choi, Hae Yun; Lee, Kwang Won; Chung, Chang Kyu

    2007-01-01

    Safety Injection Tanks (SITs) with the Fluidic Device (FD) of APR1400 provides a means of rapid reflooding of the core following a large break Loss Of Coolant Accident (LOCA), and keeping it covered until flow from the Safety Injection Pump (SIP) becomes available. A passive FD can provide two operation stages of a safety water injection into the RCS and allow more effective use of borated water in case of LOCA. Once a large break LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced to a lower flow rate. The conventional computer code 'TURTLE' used to simulate the blowdown of OPR1000 SIT can not be directly applied to simulate a blowdown process of the SIT with FD. A new computer code is needed to be developed for the blowdown test evaluation of the APR1400 SIT with FD. Korea Power Engineering Company (KOPEC) has developed a new computer code to analyze the characteristics of the SIT with FD and validated the code through the comparison of the calculation results with the test results obtained by Ulchin 5 and 6 units pre-operational test and VAlve Performance Evaluation Rig (VAPER) tests performed by The Korea Atomic Energy Research Institute (KAERI)

  5. Do Performance-Based Codes Support Universal Design in Architecture?

    DEFF Research Database (Denmark)

    Grangaard, Sidse; Frandsen, Anne Kathrine

    2016-01-01

    – Universal Design (UD). The empirical material consists of input from six workshops to which all 700 Danish Architectural firms were invited, as well as eight group interviews. The analysis shows that the current prescriptive requirements are criticized for being too homogenous and possibilities...... for differentiation and zoning are required. Therefore, a majority of professionals are interested in a performance-based model because they think that such a model will support ‘accessibility zoning’, achieving flexibility because of different levels of accessibility in a building due to its performance. The common...... of educational objectives is suggested as a tool for such a boost. The research project has been financed by the Danish Transport and Construction Agency....

  6. Performance test for a solar water heater

    Science.gov (United States)

    1979-01-01

    Two reports describe procedures and results of performance tests on domestic solar powered hot water system. Performance tests determine amount of energy collected by system, amount of energy delivered to solar source, power required to operate system and maintain proper tank temperature, overall system efficiency, and temperature distribution in tank.

  7. Optimizing fusion PIC code performance at scale on Cori Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Koskela, T. S.; Deslippe, J.

    2017-07-23

    In this paper we present the results of optimizing the performance of the gyrokinetic full-f fusion PIC code XGC1 on the Cori Phase Two Knights Landing system. The code has undergone substantial development to enable the use of vector instructions in its most expensive kernels within the NERSC Exascale Science Applications Program. We study the single-node performance of the code on an absolute scale using the roofline methodology to guide optimization efforts. We have obtained 2x speedups in single node performance due to enabling vectorization and performing memory layout optimizations. On multiple nodes, the code is shown to scale well up to 4000 nodes, near half the size of the machine. We discuss some communication bottlenecks that were identified and resolved during the work.

  8. Soft-Decision-Data Reshuffle to Mitigate Pulsed Radio Frequency Interference Impact on Low-Density-Parity-Check Code Performance

    Science.gov (United States)

    Ni, Jianjun David

    2011-01-01

    This presentation briefly discusses a research effort on mitigation techniques of pulsed radio frequency interference (RFI) on a Low-Density-Parity-Check (LDPC) code. This problem is of considerable interest in the context of providing reliable communications to the space vehicle which might suffer severe degradation due to pulsed RFI sources such as large radars. The LDPC code is one of modern forward-error-correction (FEC) codes which have the decoding performance to approach the Shannon Limit. The LDPC code studied here is the AR4JA (2048, 1024) code recommended by the Consultative Committee for Space Data Systems (CCSDS) and it has been chosen for some spacecraft design. Even though this code is designed as a powerful FEC code in the additive white Gaussian noise channel, simulation data and test results show that the performance of this LDPC decoder is severely degraded when exposed to the pulsed RFI specified in the spacecraft s transponder specifications. An analysis work (through modeling and simulation) has been conducted to evaluate the impact of the pulsed RFI and a few implemental techniques have been investigated to mitigate the pulsed RFI impact by reshuffling the soft-decision-data available at the input of the LDPC decoder. The simulation results show that the LDPC decoding performance of codeword error rate (CWER) under pulsed RFI can be improved up to four orders of magnitude through a simple soft-decision-data reshuffle scheme. This study reveals that an error floor of LDPC decoding performance appears around CWER=1E-4 when the proposed technique is applied to mitigate the pulsed RFI impact. The mechanism causing this error floor remains unknown, further investigation is necessary.

  9. Modelling Brazilian tests with FRACOD2D (FRActure propagation CODe)

    International Nuclear Information System (INIS)

    Lanaro, Flavio; Sato, Toshinori; Rinne, Mikael; Stephansson, Ove

    2008-01-01

    This study focuses on the influence of initiated cracks on the stress distribution within rock samples subjected to tensile loading by traditional Brazilian testing. The numerical analyses show that the stress distribution is only marginally affected by the considered loading boundary conditions. On the other hand, the initiation and propagation of cracks produce a stress field that is very different from that assumed by considering the rock material as continuous, homogeneous, isotropic and elastic. In the models, stress concentrations at the bridges between the cracks were found to have tensile stresses much higher than the macroscopic direct tensile strength of the intact rock. This was possible thanks to the development of large stress gradients that can be carried by the rock between the cracks. The analysis of the deformation along the sample diameter perpendicular to the loading direction might enable one to determine the macroscopic direct tensile strength of the rock or, in a real case, of the weakest grains. The strength is indicated by the point where the stress-strain curves depart from linearity. (author)

  10. Performance Testing of Download Services of COSMC

    Directory of Open Access Journals (Sweden)

    Jiří Horák

    2013-11-01

    Full Text Available The paper presents results of performance tests of download services of Czech Office of Surveying, Mapping and Cadastre according to INSPIRE  requirements. Methodology of testing is explained, including monitoring performance  of reference servers. 26 millions of random requests were generated for each monitored operation, layer and coordinate system. The temporal development of performance indicators are analyzed and discussed. Results of performance tests approve the compliance with INSPIRE qualitative requirements for download services. All monitored services satisfy requirements of latency, capacity and availability. The latency and availability requirements are fulfilled with an abundant reserve. No problems in structure and content of responses were detected.

  11. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

  12. Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1984-12-01

    The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

  13. Simulation of the fuel rod bundle test QUENCH-03 using the system codes ASTEC and ATHLET-CD

    International Nuclear Information System (INIS)

    Kruse, P.; Koch, M.K.

    2011-01-01

    The QUENCH-03 test was performed on the 21. of January 1999 at FZK (Forschungszentrum Karlsruhe) to investigate the behaviour on reflood of PWR (Pressurized Water Reactor) fuel rods with little oxidation. This paper presents the results of the simulation of QUENCH-03 performed with the version V1.3 of the integral code ASTEC (Accident Source Term Evaluation Code) which is being developed by IRSN (France) in cooperation with GRS (Germany) and with the program version 2.1A of the mechanistic code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) which is under development by GRS. At first the QUENCH test facility and the QUENCH test program in general are described. The test conduct of the test QUENCH-03 follows as well as a description of the used codes ASTEC and ATHLET-CD with the associated modeling of the test section. The results of this calculation show that during the heat-up and transient phase both codes can calculate bundle and shroud temperatures as well as the hydrogen production in good approximation to the experimental data. During the quench phase and up to the end of the test only the oxidation model PRATER of ASTEC simulates the hydrogen production very well, the other oxidation models of ASTEC cannot calculate to some extent the measured amount of hydrogen. ATHLET-CD underestimates the integral amount at the end of the test. In the ASTEC calculations the temperatures during the quench phase show qualitatively good results, only time delays on some elevations of the bundle could be noticed. ATHLET-CD reproduces the thermal behaviour up to the first temperature escalation very well, after that the temperatures are partly over-estimated. The time delay recognized in the ASTEC calculations are seen as well. The results of the integral code ASTEC emphasize that the calculation of QUENCH-03 is possible and leading to good results concerning hydrogen release and corresponding temperatures. Because the QUENCH-03 test was

  14. A Preliminary Analysis for SMART-ITL SBLOCA Tests using the MARS/KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yeon Sik; Ko, Yung Joo; Suh, Jae Seung [System Engineering and Technology Co., Ltd., Daejeon (Korea, Republic of)

    2013-10-15

    In this paper, a preliminary analysis was conducted for SMART-ITL SBLOCA tests using the MARS/KS Code. The results of this work are expected to be good guidelines for SBLOCA tests with the SMART-ITL, and used to understand the various thermal-hydraulic phenomena expected to occur in the integral-type reactor, SMART. An integral-effect test (IET) loop for SMART, SMART-ITL (or FESTA), has been designed using a volume scaling methodology. It was installed at KAERI and its commissioning tests were finished in 2012. Its height was preserved and its area and volume were scaled down to 1/49 compared with the prototype plant, SMART. The SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The objectives of IET using the SMART-ITL facility are to investigate the integral performance of the inter-connected components and possible thermal-hydraulic phenomena occurring in the SMART design, and to validate its safety for various design basis events (DBAs)

  15. A Preliminary Analysis for SMART-ITL SBLOCA Tests using the MARS/KS Code

    International Nuclear Information System (INIS)

    Cho, Yeon Sik; Ko, Yung Joo; Suh, Jae Seung

    2013-01-01

    In this paper, a preliminary analysis was conducted for SMART-ITL SBLOCA tests using the MARS/KS Code. The results of this work are expected to be good guidelines for SBLOCA tests with the SMART-ITL, and used to understand the various thermal-hydraulic phenomena expected to occur in the integral-type reactor, SMART. An integral-effect test (IET) loop for SMART, SMART-ITL (or FESTA), has been designed using a volume scaling methodology. It was installed at KAERI and its commissioning tests were finished in 2012. Its height was preserved and its area and volume were scaled down to 1/49 compared with the prototype plant, SMART. The SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The objectives of IET using the SMART-ITL facility are to investigate the integral performance of the inter-connected components and possible thermal-hydraulic phenomena occurring in the SMART design, and to validate its safety for various design basis events (DBAs)

  16. Code division multiple-access techniques in optical fiber networks. II - Systems performance analysis

    Science.gov (United States)

    Salehi, Jawad A.; Brackett, Charles A.

    1989-08-01

    A technique based on optical orthogonal codes was presented by Salehi (1989) to establish a fiber-optic code-division multiple-access (FO-CDMA) communications system. The results are used to derive the bit error rate of the proposed FO-CDMA system as a function of data rate, code length, code weight, number of users, and receiver threshold. The performance characteristics for a variety of system parameters are discussed. A means of reducing the effective multiple-access interference signal by placing an optical hard-limiter at the front end of the desired optical correlator is presented. Performance calculations are shown for the FO-CDMA with an ideal optical hard-limiter, and it is shown that using a optical hard-limiter would, in general, improve system performance.

  17. Improving performance of DS-CDMA systems using chaotic complex Bernoulli spreading codes

    Science.gov (United States)

    Farzan Sabahi, Mohammad; Dehghanfard, Ali

    2014-12-01

    The most important goal of spreading spectrum communication system is to protect communication signals against interference and exploitation of information by unintended listeners. In fact, low probability of detection and low probability of intercept are two important parameters to increase the performance of the system. In Direct Sequence Code Division Multiple Access (DS-CDMA) systems, these properties are achieved by multiplying the data information in spreading sequences. Chaotic sequences, with their particular properties, have numerous applications in constructing spreading codes. Using one-dimensional Bernoulli chaotic sequence as spreading code is proposed in literature previously. The main feature of this sequence is its negative auto-correlation at lag of 1, which with proper design, leads to increase in efficiency of the communication system based on these codes. On the other hand, employing the complex chaotic sequences as spreading sequence also has been discussed in several papers. In this paper, use of two-dimensional Bernoulli chaotic sequences is proposed as spreading codes. The performance of a multi-user synchronous and asynchronous DS-CDMA system will be evaluated by applying these sequences under Additive White Gaussian Noise (AWGN) and fading channel. Simulation results indicate improvement of the performance in comparison with conventional spreading codes like Gold codes as well as similar complex chaotic spreading sequences. Similar to one-dimensional Bernoulli chaotic sequences, the proposed sequences also have negative auto-correlation. Besides, construction of complex sequences with lower average cross-correlation is possible with the proposed method.

  18. Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

    International Nuclear Information System (INIS)

    Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

    2005-01-01

    Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

  19. Corrosion performance tests for reinforcing steel in concrete : test procedures.

    Science.gov (United States)

    2009-09-01

    The existing test method to assess the corrosion performance of reinforcing steel embedded in concrete, mainly : ASTM G109, is labor intensive, time consuming, slow to provide comparative results, and often expensive. : However, corrosion of reinforc...

  20. Input research and testing of code TOODY. Quarterly report, July--September 1971

    International Nuclear Information System (INIS)

    Haynie, G.A.

    1997-01-01

    The purpose of this report is to simplify and further explain input instructions for Code TOODY and to demonstrate the ability of the code to reproduce cylinder test results. This input is intended to be a supplement to, and not a replacement for, the existing TOODY manual. The TOODY manual should be read and understood before attempting to read this report. Problems arise in the preparation of the input data in four areas: material definition, initial shape definition, the restart feature, and the limiting of output. Aside from these areas, the code is adequately discussed in the manual, 'TOODY, A Computer Program For Calculating Problems Of Motion In Two Dimensions'

  1. Vitrification Facility integrated system performance testing report

    International Nuclear Information System (INIS)

    Elliott, D.

    1997-01-01

    This report provides a summary of component and system performance testing associated with the Vitrification Facility (VF) following construction turnover. The VF at the West Valley Demonstration Project (WVDP) was designed to convert stored radioactive waste into a stable glass form for eventual disposal in a federal repository. Following an initial Functional and Checkout Testing of Systems (FACTS) Program and subsequent conversion of test stand equipment into the final VF, a testing program was executed to demonstrate successful performance of the components, subsystems, and systems that make up the vitrification process. Systems were started up and brought on line as construction was completed, until integrated system operation could be demonstrated to produce borosilicate glass using nonradioactive waste simulant. Integrated system testing and operation culminated with a successful Operational Readiness Review (ORR) and Department of Energy (DOE) approval to initiate vitrification of high-level waste (HLW) on June 19, 1996. Performance and integrated operational test runs conducted during the test program provided a means for critical examination, observation, and evaluation of the vitrification system. Test data taken for each Test Instruction Procedure (TIP) was used to evaluate component performance against system design and acceptance criteria, while test observations were used to correct, modify, or improve system operation. This process was critical in establishing operating conditions for the entire vitrification process

  2. Selection of a computer code for Hanford low-level waste engineered-system performance assessment

    International Nuclear Information System (INIS)

    McGrail, B.P.; Mahoney, L.A.

    1995-10-01

    Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites

  3. Interpretation of Ersec tests on the backup cooling of pressurized water reactors, by using the FLIRA code

    International Nuclear Information System (INIS)

    Reviglio, Christiane

    1977-01-01

    This research thesis addresses the study of the most severe accident, or reference accident, which might occur in nuclear reactors, a clean break of a cold branch of the primary circuit, which may put the integrity of barriers against radioactive products dispersion outside of the reactor into question again. More particularly, the thesis addresses the study of the backup cooling system, and the fact that fluid flow during re-flooding must be predicted, and that heat exchange coefficients must be known in order to assess the evolution of sheath temperatures. The research comprised an experimental part which aimed at reproducing as faithfully as possible the re-flooding sequence on a tube with internal flow or on a cluster for a better core simulation. These are the ERSEC tests which are to be interpreted. It also comprised a theoretical part based on the use of computational codes which simulate the different phases of the accident and of backup fluid injection. These codes are based on physical models which describe two-phase flows and heat exchanges, and are adjusted to experimental results. The FLIRA code is used which simulates the re-flooding of a reactor duct, and determines the evolution of different values (pressure, temperatures, flow rate, and so on) during the re-flooding process. Thus, the author presents the reference accident, reports studies performed in the USA and in France (ERSEC tests), indicates the various flow regimes and describes heat exchange mechanisms during re-flooding, presents ERSEC test results, presents the FLIRA code, reports the elaboration of governing equations, indicates the various models introduced in the FLIRA code, and describes the numerical processing of equations. He finally gives a first interpretation of ERSEC tests based on the use of the FLIRA code

  4. Analysis of CSNI benchmark test on containment using the code CONTRAN

    International Nuclear Information System (INIS)

    Haware, S.K.; Ghosh, A.K.; Raj, V.V.; Kakodkar, A.

    1994-01-01

    A programme of experimental as well as analytical studies on the behaviour of nuclear reactor containment is being actively pursued. A large number ol' experiments on pressure and temperature transients have been carried out on a one-tenth scale model vapour suppression pool containment experimental facility, simulating the 220 MWe Indian Pressurised Heavy Water Reactors. A programme of development of computer codes is underway to enable prediction of containment behaviour under accident conditions. This includes codes for pressure and temperature transients, hydrogen behaviour, aerosol behaviour etc. As a part of this ongoing work, the code CONTRAN (CONtainment TRansient ANalysis) has been developed for predicting the thermal hydraulic transients in a multicompartment containment. For the assessment of the hydrogen behaviour, the models for hydrogen transportation in a multicompartment configuration and hydrogen combustion have been incorporated in the code CONTRAN. The code also has models for the heat and mass transfer due to condensation and convection heat transfer. The structural heat transfer is modeled using the one-dimensional transient heat conduction equation. Extensive validation exercises have been carried out with the code CONTRAN. The code CONTRAN has been successfully used for the analysis of the benchmark test devised by Committee on the Safety of Nuclear Installations (CSNI) of the Organisation for Economic Cooperation and Development (OECD), to test the numerical accuracy and convergence errors in the computation of mass and energy conservation for the fluid and in the computation of heat conduction in structural walls. The salient features of the code CONTRAN, description of the CSNI benchmark test and a comparison of the CONTRAN predictions with the benchmark test results are presented and discussed in the paper. (author)

  5. A code to study the water flow in a thermal test loop

    International Nuclear Information System (INIS)

    Saunier, Jean-Pierre; Duffourt, Nicole; Lago, Bernard

    1965-01-01

    A first part reports the theoretical and analytical formulation of a flow within a specific circuit used in a thermal test installation. Equations in the different parts of the circuit are developed, and their resolution for integration into a computation code is described, including boundary conditions, constants and input functions (cell characteristics, fluid characteristics, heat transfer, friction, time slicing). The second part reports an extension of this theoretical and analytical development and code development to a two-branch circuit

  6. A Study of Performance in Low-Power Tokamak Reactor with Integrated Predictive Modeling Code

    International Nuclear Information System (INIS)

    Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

    2009-07-01

    Full text: A fusion hybrid or a small fusion power output with low power tokamak reactor is presented as another useful application of nuclear fusion. Such tokamak can be used for fuel breeding, high-level waste transmutation, hydrogen production at high temperature, and testing of nuclear fusion technology components. In this work, an investigation of the plasma performance in a small fusion power output design is carried out using the BALDUR predictive integrated modeling code. The simulations of the plasma performance in this design are carried out using the empirical-based Mixed Bohm/gyro Bohm (B/gB) model, whereas the pedestal temperature model is based on magnetic and flow shear (δ α ρ ζ 2 ) stabilization pedestal width scaling. The preliminary results using this core transport model show that the central ion and electron temperatures are rather pessimistic. To improve the performance, the optimization approach are carried out by varying some parameters, such as plasma current and power auxiliary heating, which results in some improvement of plasma performance

  7. The Effects of Humor on Test Anxiety and Test Performance

    Science.gov (United States)

    Tali, Glenda

    2017-01-01

    Testing in an academic setting provokes anxiety in all students in higher education, particularly nursing students. When students experience high levels of anxiety, the resulting decline in test performance often does not represent an accurate assessment of students' academic achievement. This quantitative, experimental study examined the effects…

  8. Systemizers are better code-breakers:Self-reported systemizing predicts code-breaking performance in expert hackers and naïve participants

    Directory of Open Access Journals (Sweden)

    India eHarvey

    2016-05-01

    Full Text Available Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders - attention to detail and systemizing - may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e. crypto-analysis or code-breaking. A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001; Baron-Cohen et al., 2003. They were also tested with behavioural tasks involving code-breaking and a control task involving security x-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing was related with performance in the x-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015. We discuss the theoretical and translational implications of our findings.

  9. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    2014-08-01

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  10. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-08-15

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  11. Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

    International Nuclear Information System (INIS)

    Misfeldt, I.

    1980-01-01

    A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

  12. Integrated Performance Testing for Nonproliferation Support Project

    Energy Technology Data Exchange (ETDEWEB)

    Johns, Russell; Bultz, Garl Alan; Byers, Kenneth R.; Yaegle, William

    2013-08-20

    The objective of this workshop is to provide participants with training in testing techniques and methodologies for assessment of the performance of: Physical Protection system elements; Material Control and Accounting (MC&A) system elements.

  13. Hypervelocity Impact Test Fragment Modeling: Modifications to the Fragment Rotation Analysis and Lightcurve Code

    Science.gov (United States)

    Gouge, Michael F.

    2011-01-01

    Hypervelocity impact tests on test satellites are performed by members of the orbital debris scientific community in order to understand and typify the on-orbit collision breakup process. By analysis of these test satellite fragments, the fragment size and mass distributions are derived and incorporated into various orbital debris models. These same fragments are currently being put to new use using emerging technologies. Digital models of these fragments are created using a laser scanner. A group of computer programs referred to as the Fragment Rotation Analysis and Lightcurve code uses these digital representations in a multitude of ways that describe, measure, and model on-orbit fragments and fragment behavior. The Dynamic Rotation subroutine generates all of the possible reflected intensities from a scanned fragment as if it were observed to rotate dynamically while in orbit about the Earth. This calls an additional subroutine that graphically displays the intensities and the resulting frequency of those intensities as a range of solar phase angles in a Probability Density Function plot. This document reports the additions and modifications to the subset of the Fragment Rotation Analysis and Lightcurve concerned with the Dynamic Rotation and Probability Density Function plotting subroutines.

  14. Summary of functional and performance test procedures

    DEFF Research Database (Denmark)

    Mitzel, Jens; Gülzow, Erich; Friedrich, K. Andreas

    Different Test Modules (TM) are defined for the functional and performance characterization of a PEMFC stack. The master document TM2.00 defines requirements and methodology for parameter variation, stability and data acquisition.......Different Test Modules (TM) are defined for the functional and performance characterization of a PEMFC stack. The master document TM2.00 defines requirements and methodology for parameter variation, stability and data acquisition....

  15. Performance of Turbo Interference Cancellation Receivers in Space-Time Block Coded DS-CDMA Systems

    Directory of Open Access Journals (Sweden)

    Emmanuel Oluremi Bejide

    2008-07-01

    Full Text Available We investigate the performance of turbo interference cancellation receivers in the space time block coded (STBC direct-sequence code division multiple access (DS-CDMA system. Depending on the concatenation scheme used, we divide these receivers into the partitioned approach (PA and the iterative approach (IA receivers. The performance of both the PA and IA receivers is evaluated in Rayleigh fading channels for the uplink scenario. Numerical results show that the MMSE front-end turbo space-time iterative approach receiver (IA effectively combats the mixture of MAI and intersymbol interference (ISI. To further investigate the possible achievable data rates in the turbo interference cancellation receivers, we introduce the puncturing of the turbo code through the use of rate compatible punctured turbo codes (RCPTCs. Simulation results suggest that combining interference cancellation, turbo decoding, STBC, and RCPTC can significantly improve the achievable data rates for a synchronous DS-CDMA system for the uplink in Rayleigh flat fading channels.

  16. SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

    International Nuclear Information System (INIS)

    Dutt, D.S.; Baker, R.B.

    1975-06-01

    The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

  17. Cascade Distiller System Performance Testing Interim Results

    Science.gov (United States)

    Callahan, Michael R.; Pensinger, Stuart; Sargusingh, Miriam J.

    2014-01-01

    The Cascade Distillation System (CDS) is a rotary distillation system with potential for greater reliability and lower energy costs than existing distillation systems. Based upon the results of the 2009 distillation comparison test (DCT) and recommendations of the expert panel, the Advanced Exploration Systems (AES) Water Recovery Project (WRP) project advanced the technology by increasing reliability of the system through redesign of bearing assemblies and improved rotor dynamics. In addition, the project improved the CDS power efficiency by optimizing the thermoelectric heat pump (TeHP) and heat exchanger design. Testing at the NASA-JSC Advanced Exploration System Water Laboratory (AES Water Lab) using a prototype Cascade Distillation Subsystem (CDS) wastewater processor (Honeywell d International, Torrance, Calif.) with test support equipment and control system developed by Johnson Space Center was performed to evaluate performance of the system with the upgrades as compared to previous system performance. The system was challenged with Solution 1 from the NASA Exploration Life Support (ELS) distillation comparison testing performed in 2009. Solution 1 consisted of a mixed stream containing human-generated urine and humidity condensate. A secondary objective of this testing is to evaluate the performance of the CDS as compared to the state of the art Distillation Assembly (DA) used in the ISS Urine Processor Assembly (UPA). This was done by challenging the system with ISS analog waste streams. This paper details the results of the AES WRP CDS performance testing.

  18. Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

    2010-06-15

    Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

  19. A probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

    International Nuclear Information System (INIS)

    Bull, A.J.

    1987-01-01

    This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases. (orig./HP)

  20. Impact of the Revised Malaysian Code on Corporate Governance on Audit Committee Attributes and Firm Performance

    OpenAIRE

    KALLAMU, Basiru Salisu

    2016-01-01

    Abstract. Using a sample of 37 finance companies listed under the finance segment of Bursa Malaysia, we examined the impact of the revision to Malaysian code on corporate governance on audit committee attributes and firm performance. Our result suggests that audit committee attributes significantly improved after the Code was revised. In addition, the coefficient for audit committee and risk committee interlock has a significant negative relationship with Tobin’s Q in the period before the re...

  1. Performance Evaluation of a Novel Optimization Sequential Algorithm (SeQ Code for FTTH Network

    Directory of Open Access Journals (Sweden)

    Fazlina C.A.S.

    2017-01-01

    Full Text Available The SeQ codes has advantages, such as variable cross-correlation property at any given number of users and weights, as well as effectively suppressed the impacts of phase induced intensity noise (PIIN and multiple access interference (MAI cancellation property. The result revealed, at system performance analysis of BER = 10-09, the SeQ code capable to achieved 1 Gbps up to 60 km.

  2. Performance and complexity of tunable sparse network coding with gradual growing tuning functions over wireless networks

    OpenAIRE

    Garrido Ortiz, Pablo; Sørensen, Chres W.; Lucani Roetter, Daniel Enrique; Agüero Calvo, Ramón

    2016-01-01

    Random Linear Network Coding (RLNC) has been shown to be a technique with several benefits, in particular when applied over wireless mesh networks, since it provides robustness against packet losses. On the other hand, Tunable Sparse Network Coding (TSNC) is a promising concept, which leverages a trade-off between computational complexity and goodput. An optimal density tuning function has not been found yet, due to the lack of a closed-form expression that links density, performance and comp...

  3. Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

    Directory of Open Access Journals (Sweden)

    Bruno Gonfiotti

    2017-01-01

    Full Text Available The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0 employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

  4. Work zone performance measures pilot test.

    Science.gov (United States)

    2011-04-01

    Currently, a well-defined and validated set of metrics to use in monitoring work zone performance do not : exist. This pilot test was conducted to assist state DOTs in identifying what work zone performance : measures can and should be targeted, what...

  5. RHIC sextant test: Accelerator systems and performance

    Energy Technology Data Exchange (ETDEWEB)

    Pilat, F.; Trbojevic, D.; Ahrens, L. [and others

    1997-08-01

    One sextant of the RHIC Collider was commissioned in early 1997 with beam. We describe here the performance of the accelerator systems, instrumentation subsystems and application software. We also describe a ramping test without beam that took place after the commissioning with beam. Finally, we analyze the implications of accelerator systems performance and their impact on the planning for RHIC installation and commissioning.

  6. RHIC sextant test: Accelerator systems and performance

    International Nuclear Information System (INIS)

    Pilat, F.; Trbojevic, D.; Ahrens, L.

    1997-01-01

    One sextant of the RHIC Collider was commissioned in early 1997 with beam. We describe here the performance of the accelerator systems, instrumentation subsystems and application software. We also describe a ramping test without beam that took place after the commissioning with beam. Finally, we analyze the implications of accelerator systems performance and their impact on the planning for RHIC installation and commissioning

  7. Comparison of the ENIGMA code with experimental data on thermal performance, stable fission gas and iodine release at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

    1997-08-01

    The predictions of the ENIGMA code have been compared with data from high burn-up fuel experiments from the Halden and RISO reactors. The experiments modelled were IFA-504 and IFA-558 from Halden and the test II-5 from the RISO power burnup test series. The code has well modelled the fuel thermal performance and has provided a good measure of iodine release from pre-interlinked fuel. After interlinkage the iodine predictions remain a good fit for one experiment, but there is significant overprediction for a second experiment (IFA-558). Stable fission gas release is also well modelled and the predictions are within the expected uncertainly band throughout the burn-up range. This report presents code predictions for stable fission gas release to 40GWd/tU, iodine release measurements to 50GWd/tU and thermal performance (fuel centre temperature) to 55GWd/tU. Fuel ratings of up to 38kW/m were modelled at the high burn-up levels. The code is shown to accurately or conservatively predict all these parameters. (author). 1 ref., 6 figs.

  8. Performance of smokeless gasoline fire test facility

    International Nuclear Information System (INIS)

    Griffin, J.F.; Watkins, R.A.

    1978-01-01

    Packaging for radioactive materials must perform satisfactorily when subjected to temperatures simulating an accident involving a fire. The new thermal test facility has proved to be a reliable method for satisfactorily performing the required test. The flame provides sufficient heat to assure that the test is valid, and the temperature can be controlled satisfactorily. Also, the air and water mist systems virtually eliminate any smoke and thereby exceed the local EPA requirements. The combination of the two systems provides an inexpensive, low maintenance technique for elimination of the smoke plume

  9. Design and implementation of a software tool intended for simulation and test of real time codes

    International Nuclear Information System (INIS)

    Le Louarn, C.

    1986-09-01

    The objective of real time software testing is to show off processing errors and unobserved functional requirements or timing constraints in a code. In the perspective of safety analysis of nuclear equipments of power plants testing should be carried independently from the physical process (which is not generally available), and because casual hardware failures must be considered. We propose here a simulation and test tool, integrally software, with large interactive possibilities for testing assembly code running on microprocessor. The OST (outil d'aide a la simulation et au Test de logiciels temps reel) simulates code execution and hardware or software environment behaviour. Test execution is closely monitored and many useful informations are automatically saved. The present thesis work details, after exposing methods and tools dedicated to real time software, the OST system. We show the internal mechanisms and objects of the system: particularly ''events'' (which describe evolutions of the system under test) and mnemonics (which describe the variables). Then, we detail the interactive means available to the user for constructing the test data and the environment of the tested software. Finally, a prototype implementation is presented along with the results of the tests carried out. This demonstrates the many advantages of the use of an automatic tool over a manual investigation. As a conclusion, further developments, nececessary to complete the final tool are rewieved [fr

  10. An Examination of the Performance Based Building Code on the Design of a Commercial Building

    Directory of Open Access Journals (Sweden)

    John Greenwood

    2012-11-01

    Full Text Available The Building Code of Australia (BCA is the principal code under which building approvals in Australia are assessed. The BCA adopted performance-based solutions for building approvals in 1996. Performance-based codes are based upon a set of explicit objectives, stated in terms of a hierarchy of requirements beginning with key general objectives. With this in mind, the research presented in this paper aims to analyse the impact of the introduction of the performance-based code within Western Australia to gauge the effect and usefulness of alternative design solutions in commercial construction using a case study project. The research revealed that there are several advantages to the use of alternative designs and that all parties, in general, are in favour of the performance-based building code of Australia. It is suggested that change in the assessment process to streamline the alternative design path is needed for the greater use of the performance-based alternative. With appropriate quality control measures, minor variations to the deemed-to-satisfy provisions could easily be managed by the current and future building surveying profession.

  11. Simulation of total loss of feed water in ATLAS test facility using SPACE code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of). Central Research Inst.

    2017-08-15

    A total loss of feedwater (TLOFW) with additional failures in ATLAS test facility was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. Partial failure of the safety injection pumps (SIPs) and the pilot-operated safety relief valves (POSRVs) of pressurizer were selected as additional failures. In order to assess the capability of SPACE code, partial failure was modeled, and compared with results of OECD-ATLAS A3.1 results. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. From the results, this indicated that SPACE code has capabilities to design extension conditions, and feed and bleed operation using POSRVs and SIPs were effective for RCS cooling capability during TLOFW.

  12. Field-based tests of geochemical modeling codes: New Zealand hydrothermal systems

    International Nuclear Information System (INIS)

    Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.

    1993-12-01

    Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions

  13. Field-based tests of geochemical modeling codes usign New Zealand hydrothermal systems

    International Nuclear Information System (INIS)

    Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.

    1994-06-01

    Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions

  14. Evaluation of Design & Analysis Code, CACTUS, for Predicting Crossflow Hydrokinetic Turbine Performance

    Energy Technology Data Exchange (ETDEWEB)

    Wosnik, Martin [Univ. of New Hampshire, Durham, NH (United States). Center for Ocean Renewable Energy; Bachant, Pete [Univ. of New Hampshire, Durham, NH (United States). Center for Ocean Renewable Energy; Neary, Vincent Sinclair [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Murphy, Andrew W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-01

    CACTUS, developed by Sandia National Laboratories, is an open-source code for the design and analysis of wind and hydrokinetic turbines. While it has undergone extensive validation for both vertical axis and horizontal axis wind turbines, and it has been demonstrated to accurately predict the performance of horizontal (axial-flow) hydrokinetic turbines, its ability to predict the performance of crossflow hydrokinetic turbines has yet to be tested. The present study addresses this problem by comparing the predicted performance curves derived from CACTUS simulations of the U.S. Department of Energy’s 1:6 scale reference model crossflow turbine to those derived by experimental measurements in a tow tank using the same model turbine at the University of New Hampshire. It shows that CACTUS cannot accurately predict the performance of this crossflow turbine, raising concerns on its application to crossflow hydrokinetic turbines generally. The lack of quality data on NACA 0021 foil aerodynamic (hydrodynamic) characteristics over the wide range of angles of attack (AoA) and Reynolds numbers is identified as the main cause for poor model prediction. A comparison of several different NACA 0021 foil data sources, derived using both physical and numerical modeling experiments, indicates significant discrepancies at the high AoA experienced by foils on crossflow turbines. Users of CACTUS for crossflow hydrokinetic turbines are, therefore, advised to limit its application to higher tip speed ratios (lower AoA), and to carefully verify the reliability and accuracy of their foil data. Accurate empirical data on the aerodynamic characteristics of the foil is the greatest limitation to predicting performance for crossflow turbines with semi-empirical models like CACTUS. Future improvements of CACTUS for crossflow turbine performance prediction will require the development of accurate foil aerodynamic characteristic data sets within the appropriate ranges of Reynolds numbers and AoA.

  15. Independent assessment of TRAC and RELAP5 codes through separate effects tests

    International Nuclear Information System (INIS)

    Saha, P.; Rohatgi, U.S.; Jo, J.H.; Neymotin, L.; Slovik, G.; Yuelys-Miksis, C.; Pu, J.

    1983-01-01

    Independent assessment of TRAC-PF1 (Version 7.0), TRAC-BD1 (Version 12.0) and RELAP5/MOD1 (Cycle 14) that was initiated at BNL in FY 1982, has been completed in FY 1983. As in the previous years, emphasis at Brookhaven has been in simulating various separate-effects tests with these advanced codes and identifying the areas where further thermal-hydraulic modeling improvements are needed. The following six catetories of tests were simulated with the above codes: (1) critical flow tests (Moby-Dick nitrogen-water, BNL flashing flow, Marviken Test 24); (2) Counter-Current Flow Limiting (CCFL) tests (University of Houston, Dartmouth College single and parallel tube test); (3) level swell tests (G.E. large vessel test); (4) steam generator tests (B and W 19-tube model S.G. tests, FLECHT-SEASET U-tube S.G. tests); (5) natural circulation tests (FRIGG loop tests); and (6) post-CHF tests (Oak Ridge steady-state test)

  16. Repository seal materials performance for a SALT Repository Project 5-year code/model development plan: Draft

    International Nuclear Information System (INIS)

    1987-06-01

    This document describes an integrated laboratory testing and model development effort for the seal system for a high-level nuclear waste repository in salt. The testing and modeling efforts are designed to determine seal material response in the repository environment, to provide models of seal system components for performance assessment, and to assist in the development of seal system designs. A code/model development and performance analysis program will be performed to predict the short- and long-term response of seal materials and seal components. The results from these analyses will be used to support the material testing activities on this contract and to support performance assessment activities that are conducted in other parts of the Salt Repository Project (SRP). 48 refs., 15 figs., 4 tabs

  17. Code of practice for the release of hydrostatic test water from hydrostatic testing of petroleum liquid and gas pipelines

    International Nuclear Information System (INIS)

    1999-01-01

    This booklet describes a series of administrative procedures regarding the code of practice in Alberta for the release of hydrostatic test water from hydrostatic testing of petroleum liquid and gas pipelines. The topics covered include the registration process, the type and quality of water to use during the test, and the analytical methods to be used. Reporting schedule and record keeping information are also covered. Schedule 1 discusses the requirements for the release of hydrostatic test water to land, while Schedule 2 describes the requirements for the release of hydrostatic test water to receiving water. 3 tabs

  18. NNWSI waste form performance test development

    International Nuclear Information System (INIS)

    Bates, J.K.; Gerding, T.J.

    1984-01-01

    A test method has been developed to measure the release of radionuclides from the waste package under simulated NNWSI repository conditions, and to provide information concerning materials interactions that may occur in the repository. Data from 13 weeks of unsaturated testing are discussed and compared to that from a 13-week analog test. The data indicate that the waste form test is capable of producing consistent, reproducible results that will be useful in evaluating the role of the waste in the long-term performance of the repository. 6 references, 3 figures

  19. Description of comprehensive pump test change to ASME OM code, subsection ISTB

    International Nuclear Information System (INIS)

    Hartley, R.S.

    1994-01-01

    The American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Main Committee and Board on Nuclear Codes and Standards (BNCS) recently approved changes to ASME OM Code-1990, Subsection ISTB, Inservice Testing of Pumps in Light-Water Reactor Power Plants. The changes will be included in the 1994 addenda to ISTB. The changes, designated as the comprehensive pump test, incorporate a new, improved philosophy for testing safety-related pumps in nuclear power plants. An important philosophical difference between the open-quotes old codeclose quotes inservice testing (IST) requirements and these changes is that the changes concentrate on less frequent, more meaningful testing while minimizing damaging and uninformative low-flow testing. The comprehensive pump test change establishes a more involved biannual test for all pumps and significantly reduces the rigor of the quarterly test for standby pumps. The increased rigor and cost of the biannual comprehensive tests are offset by the reduced cost of testing and potential damage to the standby pumps, which comprise a large portion of the safety-related pumps at most plants. This paper provides background on the pump testing requirements, discusses potential industry benefits of the change, describes the development of the comprehensive pump test, and gives examples and reasons for many of the specific changes. This paper also describes additional changes to ISTB that will be included in the 1994 addenda that are associated with, but not part of, the comprehensive pump test

  20. High performance mixed optical CDMA system using ZCC code and multiband OFDM

    Science.gov (United States)

    Nawawi, N. M.; Anuar, M. S.; Junita, M. N.; Rashidi, C. B. M.

    2017-11-01

    In this paper, we have proposed a high performance network design, which is based on mixed optical Code Division Multiple Access (CDMA) system using Zero Cross Correlation (ZCC) code and multiband Orthogonal Frequency Division Multiplexing (OFDM) called catenated OFDM. In addition, we also investigate the related changing parameters such as; effective power, number of user, number of band, code length and code weight. Then we theoretically analyzed the system performance comprehensively while considering up to five OFDM bands. The feasibility of the proposed system architecture is verified via the numerical analysis. The research results demonstrated that our developed modulation solution can significantly enhanced the total number of user; improving up to 80% for five catenated bands compared to traditional optical CDMA system, with the code length equals to 80, transmitted at 622 Mbps. It is also demonstrated that the BER performance strongly depends on number of weight, especially with less number of users. As the number of weight increases, the BER performance is better.

  1. High performance mixed optical CDMA system using ZCC code and multiband OFDM

    Directory of Open Access Journals (Sweden)

    Nawawi N. M.

    2017-01-01

    Full Text Available In this paper, we have proposed a high performance network design, which is based on mixed optical Code Division Multiple Access (CDMA system using Zero Cross Correlation (ZCC code and multiband Orthogonal Frequency Division Multiplexing (OFDM called catenated OFDM. In addition, we also investigate the related changing parameters such as; effective power, number of user, number of band, code length and code weight. Then we theoretically analyzed the system performance comprehensively while considering up to five OFDM bands. The feasibility of the proposed system architecture is verified via the numerical analysis. The research results demonstrated that our developed modulation solution can significantly enhanced the total number of user; improving up to 80% for five catenated bands compared to traditional optical CDMA system, with the code length equals to 80, transmitted at 622 Mbps. It is also demonstrated that the BER performance strongly depends on number of weight, especially with less number of users. As the number of weight increases, the BER performance is better.

  2. Interpretation of the CABRI LT1 test with SAS4A-code analysis

    International Nuclear Information System (INIS)

    Sato, Ikken; Onoda, Yu-uichi

    2001-03-01

    In the CABRI-FAST LT1 test, simulating a ULOF (Unprotected Loss of Flow) accident of LMFBR, pin failure took place rather early during the transient. No fuel melting is expected at this failure because the energy injection was too low and a rapid gas-release-like response leading to coolant-channel voiding was observed. This channel voiding was followed by a gradual fuel breakup and axial relocation. With an aid of SAS4A analysis, interpretation of this test was performed. Although the original SAS4A model was not well fitted to this type of early pin failure, the global behavior after the pin failure was reasonably simulated with temporary modifications. Through this study, gas release behavior from the failed fuel pin and its effect on further transient were well understood. It was also demonstrated that the SAS4A code has a potential to simulate the post-failure behavior initiated by a very early pin failure provided that necessary model modification is given. (author)

  3. International standard problem (ISP) No. 41. Containment iodine computer code exercise based on a radioiodine test facility (RTF) experiment

    International Nuclear Information System (INIS)

    2000-04-01

    International Standard Problem (ISP) exercises are comparative exercises in which predictions of different computer codes for a given physical problem are compared with each other or with the results of a carefully controlled experimental study. The main goal of ISP exercises is to increase confidence in the validity and accuracy of the tools, which were used in assessing the safety of nuclear installations. Moreover, they enable code users to gain experience and demonstrate their competence. The ISP No. 41 exercise, computer code exercise based on a Radioiodine Test Facility (RTF) experiment on iodine behaviour in containment under severe accident conditions, is one of such ISP exercises. The ISP No. 41 exercise was borne at the recommendation at the Fourth Iodine Chemistry Workshop held at PSI, Switzerland in June 1996: 'the performance of an International Standard Problem as the basis of an in-depth comparison of the models as well as contributing to the database for validation of iodine codes'. [Proceedings NEA/CSNI/R(96)6, Summary and Conclusions NEA/CSNI/R(96)7]. COG (CANDU Owners Group), comprising AECL and the Canadian nuclear utilities, offered to make the results of a Radioiodine Test Facility (RTF) test available for such an exercise. The ISP No. 41 exercise was endorsed in turn by the FPC (PWG4's Task Group on Fission Product Phenomena in the Primary Circuit and the Containment), PWG4 (CSNI Principal Working Group on the Confinement of Accidental Radioactive Releases), and the CSNI. The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has sponsored forty-five ISP exercises over the last twenty-four years, thirteen of them in the area of severe accidents. The criteria for the selection of the RTF test as a basis for the ISP-41 exercise were; (1) complementary to other RTF tests available through the PHEBUS and ACE programmes, (2) simplicity for ease of modelling and (3) good quality data. A simple RTF experiment performed under controlled

  4. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

    International Nuclear Information System (INIS)

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files

  5. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.

  6. Reliability issues and solutions for coding social communication performance in classroom settings.

    Science.gov (United States)

    Olswang, Lesley B; Svensson, Liselotte; Coggins, Truman E; Beilinson, Jill S; Donaldson, Amy L

    2006-10-01

    To explore the utility of time-interval analysis for documenting the reliability of coding social communication performance of children in classroom settings. Of particular interest was finding a method for determining whether independent observers could reliably judge both occurrence and duration of ongoing behavioral dimensions for describing social communication performance. Four coders participated in this study. They observed and independently coded 6 social communication behavioral dimensions using handheld computers. The dimensions were mutually exclusive and accounted for all verbal and nonverbal productions during a specified time frame. The technology allowed for coding frequency and duration for each entered code. Data were collected from 20 different 2-min video segments of children in kindergarten through 3rd-grade classrooms. Data were analyzed for interobserver and intraobserver agreements using time-interval sorting and Cohen's kappa. Further, interval size and total observation length were manipulated to determine their influence on reliability. The data revealed interval sorting and kappa to be a suitable method for examining reliability of occurrence and duration of ongoing social communication behavioral dimensions. Nearly all comparisons yielded medium to large kappa values; interval size and length of observation minimally affected results. Implications The analysis procedure described in this research solves a challenge in reliability: comparing coding by independent observers of both occurrence and duration of behaviors. Results indicate the utility of a new coding taxonomy and technology for application in online observations of social communication in a classroom setting.

  7. The development of the fuel rod transient performance analysis code FTPAC

    International Nuclear Information System (INIS)

    Han Zhijie; Ji Songtao

    2014-01-01

    Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

  8. Application and Analysis of Performance of DQPSK Advanced Modulation Format in Spectral Amplitude Coding OCDMA

    Directory of Open Access Journals (Sweden)

    Abdul Latif Memon

    2014-04-01

    Full Text Available SAC (Spectral Amplitude Coding is a technique of OCDMA (Optical Code Division Multiple Access to encode and decode data bits by utilizing spectral components of the broadband source. Usually OOK (ON-Off-Keying modulation format is used in this encoding scheme. To make SAC OCDMA network spectrally efficient, advanced modulation format of DQPSK (Differential Quaternary Phase Shift Keying is applied, simulated and analyzed. m-sequence code is encoded in the simulated setup. Performance regarding various lengths of m-sequence code is also analyzed and displayed in the pictorial form. The results of the simulation are evaluated with the help of electrical constellation diagram, eye diagram and bit error rate graph. All the graphs indicate better transmission quality in case of advanced modulation format of DQPSK used in SAC OCDMA network as compared with OOK

  9. Implementation and Performance Evaluation of Distributed Cloud Storage Solutions using Random Linear Network Coding

    DEFF Research Database (Denmark)

    Fitzek, Frank; Toth, Tamas; Szabados, Áron

    2014-01-01

    This paper advocates the use of random linear network coding for storage in distributed clouds in order to reduce storage and traffic costs in dynamic settings, i.e. when adding and removing numerous storage devices/clouds on-the-fly and when the number of reachable clouds is limited. We introduce...... various network coding approaches that trade-off reliability, storage and traffic costs, and system complexity relying on probabilistic recoding for cloud regeneration. We compare these approaches with other approaches based on data replication and Reed-Solomon codes. A simulator has been developed...... to carry out a thorough performance evaluation of the various approaches when relying on different system settings, e.g., finite fields, and network/storage conditions, e.g., storage space used per cloud, limited network use, and limited recoding capabilities. In contrast to standard coding approaches, our...

  10. Typical performance of regular low-density parity-check codes over general symmetric channels

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Saad, David

    2003-01-01

    Typical performance of low-density parity-check (LDPC) codes over a general binary-input output-symmetric memoryless channel is investigated using methods of statistical mechanics. Relationship between the free energy in statistical-mechanics approach and the mutual information used in the information-theory literature is established within a general framework; Gallager and MacKay-Neal codes are studied as specific examples of LDPC codes. It is shown that basic properties of these codes known for particular channels, including their potential to saturate Shannon's bound, hold for general symmetric channels. The binary-input additive-white-Gaussian-noise channel and the binary-input Laplace channel are considered as specific channel models

  11. Typical performance of regular low-density parity-check codes over general symmetric channels

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Toshiyuki [Department of Electronics and Information Engineering, Tokyo Metropolitan University, 1-1 Minami-Osawa, Hachioji-shi, Tokyo 192-0397 (Japan); Saad, David [Neural Computing Research Group, Aston University, Aston Triangle, Birmingham B4 7ET (United Kingdom)

    2003-10-31

    Typical performance of low-density parity-check (LDPC) codes over a general binary-input output-symmetric memoryless channel is investigated using methods of statistical mechanics. Relationship between the free energy in statistical-mechanics approach and the mutual information used in the information-theory literature is established within a general framework; Gallager and MacKay-Neal codes are studied as specific examples of LDPC codes. It is shown that basic properties of these codes known for particular channels, including their potential to saturate Shannon's bound, hold for general symmetric channels. The binary-input additive-white-Gaussian-noise channel and the binary-input Laplace channel are considered as specific channel models.

  12. Application and analysis of performance of dqpsk advanced modulation format in spectral amplitude coding ocdma

    International Nuclear Information System (INIS)

    Memon, A.

    2015-01-01

    SAC (Spectral Amplitude Coding) is a technique of OCDMA (Optical Code Division Multiple Access) to encode and decode data bits by utilizing spectral components of the broadband source. Usually OOK (ON-Off-Keying) modulation format is used in this encoding scheme. To make SAC OCDMA network spectrally efficient, advanced modulation format of DQPSK (Differential Quaternary Phase Shift Keying) is applied, simulated and analyzed, m-sequence code is encoded in the simulated setup. Performance regarding various lengths of m-sequence code is also analyzed and displayed in the pictorial form. The results of the simulation are evaluated with the help of electrical constellation diagram, eye diagram and bit error rate graph. All the graphs indicate better transmission quality in case of advanced modulation format of DQPSK used in SAC OCDMA network as compared with OOK. (author)

  13. Studying the co-evolution of production and test code in open source and industrial developer test processes through repository mining

    NARCIS (Netherlands)

    Zaidman, A.; Van Rompaey, B.; Van Deursen, A.; Demeyer, S.

    2010-01-01

    Many software production processes advocate rigorous development testing alongside functional code writing, which implies that both test code and production code should co-evolve. To gain insight in the nature of this co-evolution, this paper proposes three views (realized by a tool called TeMo)

  14. Testing of a Code for the Calculation of Spectra of Neutrons Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.

  15. Pre-Test Analysis of the MEGAPIE Spallation Source Target Cooling Loop Using the TRAC/AAA Code

    International Nuclear Information System (INIS)

    Bubelis, Evaldas; Coddington, Paul; Leung, Waihung

    2006-01-01

    A pilot project is being undertaken at the Paul Scherrer Institute in Switzerland to test the feasibility of installing a Lead-Bismuth Eutectic (LBE) spallation target in the SINQ facility. Efforts are coordinated under the MEGAPIE project, the main objectives of which are to design, build, operate and decommission a 1 MW spallation neutron source. The technology and experience of building and operating a high power spallation target are of general interest in the design of an Accelerator Driven System (ADS) and in this context MEGAPIE is one of the key experiments. The target cooling is one of the important aspects of the target system design that needs to be studied in detail. Calculations were performed previously using the RELAP5/Mod 3.2.2 and ATHLET codes, but in order to verify the previous code results and to provide another capability to model LBE systems, a similar study of the MEGAPIE target cooling system has been conducted with the TRAC/AAA code. In this paper a comparison is presented for the steady-state results obtained using the above codes. Analysis of transients, such as unregulated cooling of the target, loss of heat sink, the main electro-magnetic pump trip of the LBE loop and unprotected proton beam trip, were studied with TRAC/AAA and compared to those obtained earlier using RELAP5/Mod 3.2.2. This work extends the existing validation data-base of TRAC/AAA to heavy liquid metal systems and comprises the first part of the TRAC/AAA code validation study for LBE systems based on data from the MEGAPIE test facility and corresponding inter-code comparisons. (authors)

  16. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujita, R.K.

    1985-01-01

    A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

  17. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujits, R.K.

    1985-01-01

    A computer code (TRAC-PFI/MODI; denoted as TRAC) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the Once-Through Integral Systems (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and saturation, intermittent reactor coolant system circulation, boiler-condenser mode and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool- and auxiliary- feedwater initiated boiler-condenser mode heat transfer

  18. Capability of the RELAP5 code to simulate natural circulation behaviour in test facilities

    International Nuclear Information System (INIS)

    Mangal, Amit; Jain, Vikas; Nayak, A.K.

    2011-01-01

    In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized

  19. Testing of the PELSHIE shielding code using Benchmark problems and other special shielding models

    International Nuclear Information System (INIS)

    Language, A.E.; Sartori, D.E.; De Beer, G.P.

    1981-08-01

    The PELSHIE shielding code for gamma rays from point and extended sources was written in 1971 and a revised version was published in October 1979. At Pelindaba the program is used extensively due to its flexibility and ease of use for a wide range of problems. The testing of PELSHIE results with the results of a range of models and so-called Benchmark problems is desirable to determine possible weaknesses in PELSHIE. Benchmark problems, experimental data, and shielding models, some of which were resolved by the discrete-ordinates method with the ANISN and DOT 3.5 codes, were used for the efficiency test. The description of the models followed the pattern of a classical shielding problem. After the intercomparison with six different models, the usefulness of the PELSHIE code was quantitatively determined [af

  20. Lesson learned from the application to LOBI tests of CATHARE and RELAP5 codes

    International Nuclear Information System (INIS)

    Ambrosini, W.; D'Auria, F.; Galassi, G.M.

    1992-01-01

    The Dipt. di Costruzioni Meccaniche e Nucleari has participated to the LOBI project since its very beginning, contributing to almost all the international activities in this field, such as task group meetings, International Standards Problems, Seminars, etc. System codes like RELAP4/MOD6, RELAP5/MOD1, RELAP5/MOD1-EUR, RELAP5/MOD2, CATHARE 1 and CATHARE 2 were applied to the design and post test evaluation of a wide series of both LOBI/MOD1 and LOBI/MOD2 experiments, including Large Break LOCAs, Small and Intermediate Break LOCAs, long lasting transients and characterization tests. The LOBI data base demonstrated its usefulness in assessing capabilities and limitations of these codes and in qualifying a code use strategy. (author)

  1. Alternate performance standard project: Interpreting the post-construction test

    International Nuclear Information System (INIS)

    Williamson, A.D.; McDonough, S.E.

    1993-01-01

    The paper describes the results of a project commissioned by the State of Florida, in cooperation with the US Environmental Protection Agency, as one portion of the Florida Radon Research Program (FRRP). The purpose of the FRRP is to provide technical support for a statewide Building Standard for Radon-Resistant Construction currently in the rulemaking process. In this case the information provides technical background for a post-construction radon test specified as a performance element of the code which accompanies the prescriptive alternative that does not incorporate active radon reduction systems

  2. NRC valve performance test program - check valve testing

    International Nuclear Information System (INIS)

    Jeanmougin, N.M.

    1987-01-01

    The Valve Performance Test Program addresses the current requirements for testing of pressure isolation valves (PIVs) in light water reactors. Leak rate monitoring is the current method used by operating commercial power plants to survey the condition of their PIVs. ETEC testing of three check valves (4-inch, 6-inch, and 12-inch nominal diameters) indicates that leak rate testing is not a reliable method for detecting impending valve failure. Acoustic emission monitoring of check valves shows promise as a method of detecting loosened internals damage. Future efforts will focus on evaluation of acoustic emission monitoring as a technique for determining check valve condition. Three gate valves also will be tested to evaluate whether the check valve results are applicable to gate type PIVs

  3. Performance testing of extremity dosimeters, Study 2

    International Nuclear Information System (INIS)

    Harty, R.; Reece, W.D.; Hooker, C.D.

    1990-04-01

    The Health Physics Society Standards Committee (HPSSC) Working Group on Performance Testing of Extremity Dosimeters has issued a draft of a proposed standard for extremity dosimeters. The draft standard proposes methods to be used for testing dosimetry systems that determine occupational radiation dose to the extremities and the performance criterion used to determine compliance with the standard. Pacific Northwest Laboratory (PNL) has conducted two separate evaluations of the performance of extremity dosimeter processors to determine the appropriateness of the draft standard, as well as to obtain information regarding the performance of extremity dosimeters. Based on the information obtained during the facility visits and the results obtained from the performance testing, it was recommended that changes be made to ensure that the draft standard is appropriate for extremity dosimeters. The changes include: subdividing the mixture category and the beta particle category; eliminating the neutron category until appropriate flux-to-dose equivalent conversion factors are derived; and changing the tolerance level for the performance criterion to provide consistency with the performance criterion for whole body dosimeters, and to avoid making the draft standard overly difficult for processors of extremity dosimeters to pass. 20 refs., 10 figs., 6 tabs

  4. PORST: a computer code to analyze the performance of retrofitted steam turbines

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C.; Hwang, I.T.

    1980-09-01

    The computer code PORST was developed to analyze the performance of a retrofitted steam turbine that is converted from a single generating to a cogenerating unit for purposes of district heating. Two retrofit schemes are considered: one converts a condensing turbine to a backpressure unit; the other allows the crossover extraction of steam between turbine cylinders. The code can analyze the performance of a turbine operating at: (1) valve-wide-open condition before retrofit, (2) partial load before retrofit, (3) valve-wide-open after retrofit, and (4) partial load after retrofit.

  5. 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...

  6. Performance Analysis of Spectral Amplitude Coding Based OCDMA System with Gain and Splitter Mismatch

    Science.gov (United States)

    Umrani, Fahim A.; Umrani, A. Waheed; Umrani, Naveed A.; Memon, Kehkashan A.; Kalwar, Imtiaz Hussain

    2013-09-01

    This paper presents the practical analysis of the optical code-division multiple-access (O-CDMA) systems based on perfect difference codes. The work carried out use SNR criterion to select the optimal value of avalanche photodiodes (APD) gain and shows how the mismatch in the splitters and gains of the APD used in the transmitters and receivers of network can degrade the BER performance of the system. The investigations also reveal that higher APD gains are not suitable for such systems even at higher powers. The system performance, with consideration of shot noise, thermal noise, bulk and surface leakage currents is also investigated.

  7. Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ

    International Nuclear Information System (INIS)

    Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2014-08-01

    In this work was realized the test and validation of an iterative code for neutronic spectrometry known as Neutron Spectrometry and Dosimetry of the Universidad Autonoma de Zacatecas (NSDUAZ). This code was designed in a user graph interface, friendly and intuitive in the environment programming of LabVIEW using the iterative algorithm known as SPUNIT. The main characteristics of the program are: the automatic selection of the initial spectrum starting from the neutrons spectra catalog compiled by the International Atomic Energy Agency, the possibility to generate a report in HTML format that shows in graph and numeric way the neutrons flowing and calculates the ambient dose equivalent with base to this. To prove the designed code, the count rates of a spectrometer system of Bonner spheres were used with a detector of 6 LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: 252 Cf and 239 PuBe were used to validate the code, the obtained results were compared against those obtained using the BUNKIUT code. We find that the reconstructed spectra present an error that is inside the limit reported in the literature that oscillates around 15%. Therefore, it was concluded that the designed code presents similar results to those techniques used at the present time. (Author)

  8. Development of turbopump cavitation performance test facility and the test of inducer performance

    International Nuclear Information System (INIS)

    Sohn, Dong Kee; Kim, Chun Tak; Yoon, Min Soo; Cha, Bong Jun; Kim, Jin Han; Yang, Soo Seok

    2001-01-01

    A performance test facility for turbopump inducer cavitation was developed and the inducer cavitation performance tests were performed. Major components of the performance test facility are driving unit, test section, piping, water tank, and data acquisition and control system. The maximum of testing capability of this facility are as follows: flow rate - 30kg/s; pressure - 13 bar, rotational speed - 10,000rpm. This cavitation test facility is characterized by the booster pump installed at the outlet of the pump that extends the flow rate range, and by the pressure control system that makes the line pressure down to vapor pressure. The vacuum pump is used for removing the dissolved air in the water as well as the line pressure. Performance tests were carried out and preliminary data of test model inducer were obtained. The cavitation performance test and cavitation bubble flow visualization were also made. This facility is originally designed for turbopump inducer performance test and cavitation test. However it can be applied to the pump impeller performance test in the future with little modification

  9. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  10. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  11. Using Association Rules to Study the Co-evolution of Production & Test Code

    NARCIS (Netherlands)

    Lubsen, Z.; Zaidman, A.; Pinzger, M.

    2009-01-01

    Paper accepted for publication in the proceedings of the 6th International Working Conference on Mining Software Repositories (MSR 2009). Unit tests are generally acknowledged as an important aid to produce high quality code, as they provide quick feedback to developers on the correctness of their

  12. Studying Co-evolution of Production and Test Code Using Association Rule Mining

    NARCIS (Netherlands)

    Lubsen, Z.; Zaidman, A.; Pinzger, M.

    2009-01-01

    Long version of the short paper accepted for publication in the proceedings of the 6th International Working Conference on Mining Software Repositories (MSR 2009). Unit tests are generally acknowledged as an important aid to produce high quality code, as they provide quick feedback to developers on

  13. User input verification and test driven development in the NJOY21 nuclear data processing code

    Energy Technology Data Exchange (ETDEWEB)

    Trainer, Amelia Jo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCartney, Austin Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-21

    Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, and capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.

  14. Testing prospect theory in students’ performance

    OpenAIRE

    Pérez Galdón, Patricia; Nicolau, Juan Luis

    2013-01-01

    This paper tests the existence of ‘reference dependence’ and ‘loss aversion’ in students’ academic performance. Accordingly, achieving a worse than expected academic performance would have a much stronger effect on students’ (dis)satisfaction than obtaining a better than expected grade. Although loss aversion is a well-established finding, some authors have demonstrated that it can be moderated – diminished, to be precise–. Within this line of research, we also examine whether the students’ e...

  15. Load responsive multilayer insulation performance testing

    International Nuclear Information System (INIS)

    Dye, S.; Kopelove, A.; Mills, G. L.

    2014-01-01

    Cryogenic insulation designed to operate at various pressures from one atmosphere to vacuum, with high thermal performance and light weight, is needed for cryogenically fueled space launch vehicles and aircraft. Multilayer insulation (MLI) performs well in a high vacuum, but the required vacuum shell for use in the atmosphere is heavy. Spray-on foam insulation (SOFI) is often used in these systems because of its light weight, but can have a higher heat flux than desired. We report on the continued development of Load Responsive Multilayer Insulation (LRMLI), an advanced thermal insulation system that uses dynamic beam discrete spacers that provide high thermal performance both in atmosphere and vacuum. LRMLI consists of layers of thermal radiation barriers separated and supported by micromolded polymer spacers. The spacers have low thermal conductance, and self-support a thin, lightweight vacuum shell that provides internal high vacuum in the insulation. The dynamic load responsive spacers compress to support the external load of a vacuum shell in one atmosphere, and decompress under reduced atmospheric pressure for lower heat leak. Structural load testing was performed on the spacers with various configurations. LRMLI was installed on a 400 liter tank and boil off testing with liquid nitrogen performed at various chamber pressures from one atmosphere to high vacuum. Testing was also performed with an MLI blanket on the outside of the LRMLI

  16. Load responsive multilayer insulation performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Dye, S.; Kopelove, A. [Quest Thermal Group, 6452 Fig Street Suite A, Arvada, CO 80004 (United States); Mills, G. L. [Ball Aerospace and Technologies Corp, 1600 Commerce Street, Boulder, CO 80301 (United States)

    2014-01-29

    Cryogenic insulation designed to operate at various pressures from one atmosphere to vacuum, with high thermal performance and light weight, is needed for cryogenically fueled space launch vehicles and aircraft. Multilayer insulation (MLI) performs well in a high vacuum, but the required vacuum shell for use in the atmosphere is heavy. Spray-on foam insulation (SOFI) is often used in these systems because of its light weight, but can have a higher heat flux than desired. We report on the continued development of Load Responsive Multilayer Insulation (LRMLI), an advanced thermal insulation system that uses dynamic beam discrete spacers that provide high thermal performance both in atmosphere and vacuum. LRMLI consists of layers of thermal radiation barriers separated and supported by micromolded polymer spacers. The spacers have low thermal conductance, and self-support a thin, lightweight vacuum shell that provides internal high vacuum in the insulation. The dynamic load responsive spacers compress to support the external load of a vacuum shell in one atmosphere, and decompress under reduced atmospheric pressure for lower heat leak. Structural load testing was performed on the spacers with various configurations. LRMLI was installed on a 400 liter tank and boil off testing with liquid nitrogen performed at various chamber pressures from one atmosphere to high vacuum. Testing was also performed with an MLI blanket on the outside of the LRMLI.

  17. Review of FRAP-T4 performance based on fuel behavior tests conducted in the PBF

    International Nuclear Information System (INIS)

    Charyulu, M.K.

    1979-09-01

    The ability of the Fuel Rod Analysis Program - Transient (FRAP-T), a computer code developed at the Idaho National Engineering Laboratory to calculate fuel rod behavior during transient experiments conducted in the Power Burst Facility, is discussed. Fuel rod behavior calculations are compared with data from tests performed under postulated RIA, LOCA, and PCM accident conditions. Physical phenomena, rod damage, and damage mechanisms observed during the tests and not presently incorporated into the FRAP-T code are identified

  18. Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  19. Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  20. Application of advanced validation concepts to oxide fuel performance codes: LIFE-4 fast-reactor and FRAPCON thermal-reactor fuel performance codes

    Energy Technology Data Exchange (ETDEWEB)

    Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-10-15

    Highlights: ► The application of advanced validation techniques (sensitivity, calibration and prediction) to nuclear performance codes FRAPCON and LIFE-4 is the focus of the paper. ► A sensitivity ranking methodology narrows down the number of selected modeling parameters from 61 to 24 for FRAPCON and from 69 to 35 for LIFE-4. ► Fuel creep, fuel thermal conductivity, fission gas transport/release, crack/boundary, and fuel gap conductivity models of LIFE-4 are identified for improvements. ► FRAPCON sensitivity results indicated the importance of the fuel thermal conduction and the fission gas release models. -- Abstract: Evolving nuclear energy programs expect to use enhanced modeling and simulation (M and S) capabilities, using multiscale, multiphysics modeling approaches, to reduce both cost and time from the design through the licensing phases. Interest in the development of the multiscale, multiphysics approach has increased in the last decade because of the need for predictive tools for complex interacting processes as a means of eliminating the limited use of empirically based model development. Complex interacting processes cannot be predicted by analyzing each individual component in isolation. In most cases, the mathematical models of complex processes and their boundary conditions are nonlinear. As a result, the solutions of these mathematical models often require high-performance computing capabilities and resources. The use of multiscale, multiphysics (MS/MP) models in conjunction with high-performance computational software and hardware introduces challenges in validating these predictive tools—traditional methodologies will have to be modified to address these challenges. The advanced MS/MP codes for nuclear fuels and reactors are being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the US Department of Energy (DOE) – Nuclear Energy (NE). This paper does not directly address challenges in calibration

  1. MILC Code Performance on High End CPU and GPU Supercomputer Clusters

    Science.gov (United States)

    DeTar, Carleton; Gottlieb, Steven; Li, Ruizi; Toussaint, Doug

    2018-03-01

    With recent developments in parallel supercomputing architecture, many core, multi-core, and GPU processors are now commonplace, resulting in more levels of parallelism, memory hierarchy, and programming complexity. It has been necessary to adapt the MILC code to these new processors starting with NVIDIA GPUs, and more recently, the Intel Xeon Phi processors. We report on our efforts to port and optimize our code for the Intel Knights Landing architecture. We consider performance of the MILC code with MPI and OpenMP, and optimizations with QOPQDP and QPhiX. For the latter approach, we concentrate on the staggered conjugate gradient and gauge force. We also consider performance on recent NVIDIA GPUs using the QUDA library.

  2. MILC Code Performance on High End CPU and GPU Supercomputer Clusters

    Directory of Open Access Journals (Sweden)

    DeTar Carleton

    2018-01-01

    Full Text Available With recent developments in parallel supercomputing architecture, many core, multi-core, and GPU processors are now commonplace, resulting in more levels of parallelism, memory hierarchy, and programming complexity. It has been necessary to adapt the MILC code to these new processors starting with NVIDIA GPUs, and more recently, the Intel Xeon Phi processors. We report on our efforts to port and optimize our code for the Intel Knights Landing architecture. We consider performance of the MILC code with MPI and OpenMP, and optimizations with QOPQDP and QPhiX. For the latter approach, we concentrate on the staggered conjugate gradient and gauge force. We also consider performance on recent NVIDIA GPUs using the QUDA library.

  3. Improving 3D-Turbo Code's BER Performance with a BICM System over Rayleigh Fading Channel

    Directory of Open Access Journals (Sweden)

    R. Yao

    2016-12-01

    Full Text Available Classical Turbo code suffers from high error floor due to its small Minimum Hamming Distance (MHD. Newly-proposed 3D-Turbo code can effectively increase the MHD and achieve a lower error floor by adding a rate-1 post encoder. In 3D-Turbo codes, part of the parity bits from the classical Turbo encoder are further encoded through the post encoder. In this paper, a novel Bit-Interleaved Coded Modulation (BICM system is proposed by combining rotated mapping Quadrature Amplitude Modulation (QAM and 3D-Turbo code to improve the Bit Error Rate (BER performance of 3D-Turbo code over Raleigh fading channel. A key-bit protection scheme and a Two-Dimension (2D iterative soft demodulating-decoding algorithm are developed for the proposed BICM system. Simulation results show that the proposed system can obtain about 0.8-1.0 dB gain at BER of 10^{-6}, compared with the existing BICM system with Gray mapping QAM.

  4. NetBench. Automated Network Performance Testing

    CERN Document Server

    Cadeddu, Mattia

    2016-01-01

    In order to evaluate the operation of high performance routers, CERN has developed the NetBench software to run benchmarking tests by injecting various traffic patterns and observing the network devices behaviour in real-time. The tool features a modular design with a Python based console used to inject traffic and collect the results in a database, and a web user

  5. Testing for Distortions in Performance Measures

    DEFF Research Database (Denmark)

    Sloof, Randolph; Van Praag, Mirjam

    2015-01-01

    Distorted performance measures in compensation contracts elicit suboptimal behavioral responses that may even prove to be dysfunctional (gaming). This paper applies the empirical test developed by Courty and Marschke (Review of Economics and Statistics, 90, 428-441) to detect whether the widely...

  6. Testing for Distortions in Performance Measures

    DEFF Research Database (Denmark)

    Sloof, Randolph; Van Praag, Mirjam

    Distorted performance measures in compensation contracts elicit suboptimal behavioral responses that may even prove to be dysfunctional (gaming). This paper applies the empirical test developed by Courty and Marschke (2008) to detect whether the widely used class of Residual Income based performa...

  7. Drop Performance Test of CRDMs for JRTR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myoung-Hwan; Cho, Yeong-Garp; Chung, Jong-Ha [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Jung-Hyun [POSCO Plandtec Co. Ltd, Ulsan (Korea, Republic of); Lee, Kwan-Hee [RIST, Pohang (Korea, Republic of)

    2015-10-15

    The drop test results of CRDMs with AC-type electromagnet show that the initial delay times are not satisfied with the requirement, 0.15 seconds. After the replacement of the electromagnet from AC-type to DCtype, the drop times of CARs and accelerations due to the impact of moving parts are satisfied with all requirements. As a result, it is found that four CRDMs to be installed at site have a good drop performance, and meet all performance requirements. A control rod drive mechanism (CRDM) is a device to control the position of a control absorber rod (CAR) in the core by using a stepping motor which is commanded by the reactor regulating system (RRS) to control the reactivity during the normal operation of the reactor. The top-mounted CRDM driven by the stepping motor for Jordan Research and Training Reactor (JRTR) has been developed in KAERI. The CRDM for JRTR has been optimized by the design improvement based on that of the HANARO. It is necessary to verify the performances such as the stepping, drop, endurance, vibration, seismic and structural integrity for active components. Especially, the CAR drop curves are important data for the safety analysis. This paper describes the test results to demonstrate the drop performances of a prototype and 4 CRDMs to be installed at site. The tests are carried out at a test rig simulating the actual reactor's conditions.

  8. SIMS Prototype System 4: performance test report

    Energy Technology Data Exchange (ETDEWEB)

    1978-10-09

    The results obtained during testing of a self-contained, preassembled air type solar system, designed for installation remote from the dwelling, to provide space heating and hot water are presented. Data analysis is included which documents the system performance and verifies the suitability of SIMS Prototype System 4 for field installation.

  9. Introduction to the Latest Version of the Test-Particle Monte Carlo Code Molflow+

    CERN Document Server

    Ady, M

    2014-01-01

    The Test-Particle Monte Carlo code Molflow+ is getting more and more attention from the scientific community needing detailed 3D calculations of vacuum in the molecular flow regime mainly, but not limited to, the particle accelerator field. Substantial changes, bug fixes, geometry-editing and modelling features, and computational speed improvements have been made to the code in the last couple of years. This paper will outline some of these new features, and show examples of applications to the design and analysis of vacuum systems at CERN and elsewhere.

  10. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  11. A Test Data Compression Scheme Based on Irrational Numbers Stored Coding

    Directory of Open Access Journals (Sweden)

    Hai-feng Wu

    2014-01-01

    Full Text Available Test question has already become an important factor to restrict the development of integrated circuit industry. A new test data compression scheme, namely irrational numbers stored (INS, is presented. To achieve the goal of compress test data efficiently, test data is converted into floating-point numbers, stored in the form of irrational numbers. The algorithm of converting floating-point number to irrational number precisely is given. Experimental results for some ISCAS 89 benchmarks show that the compression effect of proposed scheme is better than the coding methods such as FDR, AARLC, INDC, FAVLC, and VRL.

  12. A test data compression scheme based on irrational numbers stored coding.

    Science.gov (United States)

    Wu, Hai-feng; Cheng, Yu-sheng; Zhan, Wen-fa; Cheng, Yi-fei; Wu, Qiong; Zhu, Shi-juan

    2014-01-01

    Test question has already become an important factor to restrict the development of integrated circuit industry. A new test data compression scheme, namely irrational numbers stored (INS), is presented. To achieve the goal of compress test data efficiently, test data is converted into floating-point numbers, stored in the form of irrational numbers. The algorithm of converting floating-point number to irrational number precisely is given. Experimental results for some ISCAS 89 benchmarks show that the compression effect of proposed scheme is better than the coding methods such as FDR, AARLC, INDC, FAVLC, and VRL.

  13. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP-CENTERAL RESEARCH INSTITUTE, Daejeon (Korea, Republic of)

    2013-10-15

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic

  14. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    International Nuclear Information System (INIS)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

    2013-01-01

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic state in

  15. Modification in the FUDA computer code to predict fuel performance at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Das, M; Arunakumar, B V; Prasad, P N [Nuclear Power Corp., Mumbai (India)

    1997-08-01

    The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig.

  16. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  17. Modification in the FUDA computer code to predict fuel performance at high burnup

    International Nuclear Information System (INIS)

    Das, M.; Arunakumar, B.V.; Prasad, P.N.

    1997-01-01

    The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig

  18. Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

    International Nuclear Information System (INIS)

    Edelmann, Paul Guy; Williams, Brian J.; Unal, Cetin; Yacout, Abdellatif

    2012-01-01

    This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

  19. Motivation and Test Anxiety in Test Performance across Three Testing Contexts: The CAEL, CET, and GEPT

    Science.gov (United States)

    Cheng, Liying; Klinger, Don; Fox, Janna; Doe, Christine; Jin, Yan; Wu, Jessica

    2014-01-01

    This study examined test-takers' motivation, test anxiety, and test performance across a range of social and educational contexts in three high-stakes language tests: the Canadian Academic English Language (CAEL) Assessment in Canada, the College English Test (CET) in the People's Republic of China, and the General English Proficiency Test (GEPT)…

  20. Performance analysis of a decoding algorithm for algebraic-geometry codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Jensen, Helge Elbrønd; Nielsen, Rasmus Refslund

    1999-01-01

    The fast decoding algorithm for one point algebraic-geometry codes of Sakata, Elbrond Jensen, and Hoholdt corrects all error patterns of weight less than half the Feng-Rao minimum distance. In this correspondence we analyze the performance of the algorithm for heavier error patterns. It turns out...

  1. Reliability in the performance-based concept of fib Model Code 2010

    NARCIS (Netherlands)

    Bigaj-van Vliet, A.; Vrouwenvelder, T.

    2013-01-01

    The design philosophy of the new fib Model Code for Concrete Structures 2010 represents the state of the art with regard to performance-based approach to the design and assessment of concrete structures. Given the random nature of quantities determining structural behaviour, the assessment of

  2. User's manual for the vertical axis wind turbine performance computer code darter

    Energy Technology Data Exchange (ETDEWEB)

    Klimas, P. C.; French, R. E.

    1980-05-01

    The computer code DARTER (DARrieus, Turbine, Elemental Reynolds number) is an aerodynamic performance/loads prediction scheme based upon the conservation of momentum principle. It is the latest evolution in a sequence which began with a model developed by Templin of NRC, Canada and progressed through the Sandia National Laboratories-developed SIMOSS (SSImple MOmentum, Single Streamtube) and DART (SARrieus Turbine) to DARTER.

  3. Performance of super-orthogonal space-time trellis code in a multipath environment

    CSIR Research Space (South Africa)

    Sokoya, OA

    2007-09-01

    Full Text Available This paper investigates the performance of Super-Orthogonal Space-time Trellis Code (SOSTTC) designed primarily for non-frequency selective (i.e. flat) fading channel but now applied to a frequency selective fading channel. A new decoding trellis...

  4. An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

    International Nuclear Information System (INIS)

    Hall, D.G.; Watkins, J.C.

    1987-01-01

    This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

  5. Development and feasibility testing of the Pediatric Emergency Discharge Interaction Coding Scheme.

    Science.gov (United States)

    Curran, Janet A; Taylor, Alexandra; Chorney, Jill; Porter, Stephen; Murphy, Andrea; MacPhee, Shannon; Bishop, Andrea; Haworth, Rebecca

    2017-08-01

    Discharge communication is an important aspect of high-quality emergency care. This study addresses the gap in knowledge on how to describe discharge communication in a paediatric emergency department (ED). The objective of this feasibility study was to develop and test a coding scheme to characterize discharge communication between health-care providers (HCPs) and caregivers who visit the ED with their children. The Pediatric Emergency Discharge Interaction Coding Scheme (PEDICS) and coding manual were developed following a review of the literature and an iterative refinement process involving HCP observations, inter-rater assessments and team consensus. The coding scheme was pilot-tested through observations of HCPs across a range of shifts in one urban paediatric ED. Overall, 329 patient observations were carried out across 50 observational shifts. Inter-rater reliability was evaluated in 16% of the observations. The final version of the PEDICS contained 41 communication elements. Kappa scores were greater than .60 for the majority of communication elements. The most frequently observed communication elements were under the Introduction node and the least frequently observed were under the Social Concerns node. HCPs initiated the majority of the communication. Pediatric Emergency Discharge Interaction Coding Scheme addresses an important gap in the discharge communication literature. The tool is useful for mapping patterns of discharge communication between HCPs and caregivers. Results from our pilot test identified deficits in specific areas of discharge communication that could impact adherence to discharge instructions. The PEDICS would benefit from further testing with a different sample of HCPs. © 2017 The Authors. Health Expectations Published by John Wiley & Sons Ltd.

  6. MFTF test coil construction and performance

    International Nuclear Information System (INIS)

    Cornish, D.N.; Zbasnik, J.P.; Leber, R.L.; Hirzel, D.G.; Johnston, J.E.; Rosdahl, A.R.

    1978-01-01

    A solenoid coil, 105 cm inside the 167 cm outside diameter, has been constructed and tested to study the performance of the stabilized Nb--Ti conductor to be used in the Mirror Fusion Test Facility (MFTF) being built at Lawrence Livermore Laboratory. The insulation system of the test coil is identical to that envisioned for MFTF. Cold-weld joints were made in the conductor at the start and finish of each layer; heaters were fitted to some of these joints and also to the conductor at various locations in the winding. This paper gives details of the construction of the coil and the results of the tests carried out to determine its propagation and recovery characteristics

  7. Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon-Ho; Kim, Dong-Min; Park, Chang-Hwan [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.

  8. Large-scale, multi-compartment tests in PANDA for LWR-containment analysis and code validation

    International Nuclear Information System (INIS)

    Paladino, Domenico; Auban, Olivier; Zboray, Robert

    2006-01-01

    The large-scale thermal-hydraulic PANDA facility has been used for the last years for investigating passive decay heat removal systems and related containment phenomena relevant for next-generation and current light water reactors. As part of the 5. EURATOM framework program project TEMPEST, a series of tests was performed in PANDA to experimentally investigate the distribution of hydrogen inside the containment and its effect on the performance of the Passive Containment Cooling System (PCCS) designed for the Economic Simplified Boiling Water Reactor (ESBWR). In a postulated severe accident, a large amount of hydrogen could be released in the Reactor Pressure Vessel (RPV) as a consequence of the cladding Metal- Water (M-W) reaction and discharged together with steam to the Drywell (DW) compartment. In PANDA tests, hydrogen was simulated by using helium. This paper illustrates the results of a TEMPEST test performed in PANDA and named as Test T1.2. In Test T1.2, the gas stratification (steam-helium) patterns forming in the large-scale multi-compartment PANDA DW, and the effect of non-condensable gas (helium) on the overall behaviour of the PCCS were identified. Gas mixing and stratification in a large-scale multi-compartment system are currently being further investigated in PANDA in the frame of the OECD project SETH. The testing philosophy in this new PANDA program is to produce data for code validation in relation to specific phenomena, such as: gas stratification in the containment, gas transport between containment compartments, wall condensation, etc. These types of phenomena are driven by buoyant high-momentum injections (jets) and/or low momentum injection (plumes), depending on the transient scenario. In this context, the new SETH tests in PANDA are particularly valuable to produce an experimental database for code assessment. This paper also presents an overview of the PANDA SETH tests and the major improvements in instrumentation carried out in the PANDA

  9. Performance of asynchronous fiber-optic code division multiple access system based on three-dimensional wavelength/time/space codes and its link analysis.

    Science.gov (United States)

    Singh, Jaswinder

    2010-03-10

    A novel family of three-dimensional (3-D) wavelength/time/space codes for asynchronous optical code-division-multiple-access (CDMA) systems with "zero" off-peak autocorrelation and "unity" cross correlation is reported. Antipodal signaling and differential detection is employed in the system. A maximum of [(W x T+1) x W] codes are generated for unity cross correlation, where W and T are the number of wavelengths and time chips used in the code and are prime. The conditions for violation of the cross-correlation constraint are discussed. The expressions for number of generated codes are determined for various code dimensions. It is found that the maximum number of codes are generated for S systems. The codes have a code-set-size to code-size ratio greater than W/S. For instance, with a code size of 2065 (59 x 7 x 5), a total of 12,213 users can be supported, and 130 simultaneous users at a bit-error rate (BER) of 10(-9). An arrayed-waveguide-grating-based reconfigurable encoder/decoder design for 2-D implementation for the 3-D codes is presented so that the need for multiple star couplers and fiber ribbons is eliminated. The hardware requirements of the coders used for various modulation/detection schemes are given. The effect of insertion loss in the coders is shown to be significantly reduced with loss compensation by using an amplifier after encoding. An optical CDMA system for four users is simulated and the results presented show the improvement in performance with the use of loss compensation.

  10. Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-05-01

    One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

  11. System Performance of Concatenated STBC and Block Turbo Codes in Dispersive Fading Channels

    Directory of Open Access Journals (Sweden)

    Kam Tai Chan

    2005-05-01

    Full Text Available A new scheme of concatenating the block turbo code (BTC with the space-time block code (STBC for an OFDM system in dispersive fading channels is investigated in this paper. The good error correcting capability of BTC and the large diversity gain characteristics of STBC can be achieved simultaneously. The resulting receiver outperforms the iterative convolutional Turbo receiver with maximum- a-posteriori-probability expectation maximization (MAP-EM algorithm. Because of its ability to perform the encoding and decoding processes in parallel, the proposed system is easy to implement in real time.

  12. Performance of an Error Control System with Turbo Codes in Powerline Communications

    Directory of Open Access Journals (Sweden)

    Balbuena-Campuzano Carlos Alberto

    2014-07-01

    Full Text Available This paper reports the performance of turbo codes as an error control technique in PLC (Powerline Communications data transmissions. For this system, computer simulations are used for modeling data networks based on the model classified in technical literature as indoor, and uses OFDM (Orthogonal Frequency Division Multiplexing as a modulation technique. Taking into account the channel, modulation and turbo codes, we propose a methodology to minimize the bit error rate (BER, as a function of the average received signal noise ratio (SNR.

  13. Performance and Complexity of Tunable Sparse Network Coding with Gradual Growing Tuning Functions over Wireless Networks

    DEFF Research Database (Denmark)

    Garrido, Pablo; Sørensen, Chres Wiant; Roetter, Daniel Enrique Lucani

    2016-01-01

    Random Linear Network Coding (RLNC) has been shown to be a technique with several benefits, in particular when applied over wireless mesh networks, since it provides robustness against packet losses. On the other hand, Tunable Sparse Network Coding (TSNC) is a promising concept, which leverages...... a trade-off between computational complexity and goodput. An optimal density tuning function has not been found yet, due to the lack of a closed-form expression that links density, performance and computational cost. In addition, it would be difficult to implement, due to the feedback delay. In this work...

  14. Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

    Directory of Open Access Journals (Sweden)

    Ahmed Amin El Said Abd El Hameed

    2015-08-01

    Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

  15. Using clinical data to predict high-cost performance coding issues associated with pressure ulcers: a multilevel cohort model.

    Science.gov (United States)

    Padula, William V; Gibbons, Robert D; Pronovost, Peter J; Hedeker, Donald; Mishra, Manish K; Makic, Mary Beth F; Bridges, John Fp; Wald, Heidi L; Valuck, Robert J; Ginensky, Adam J; Ursitti, Anthony; Venable, Laura Ruth; Epstein, Ziv; Meltzer, David O

    2017-04-01

    Hospital-acquired pressure ulcers (HAPUs) have a mortality rate of 11.6%, are costly to treat, and result in Medicare reimbursement penalties. Medicare codes HAPUs according to Agency for Healthcare Research and Quality Patient-Safety Indicator 3 (PSI-03), but they are sometimes inappropriately coded. The objective is to use electronic health records to predict pressure ulcers and to identify coding issues leading to penalties. We evaluated all hospitalized patient electronic medical records at an academic medical center data repository between 2011 and 2014. These data contained patient encounter level demographic variables, diagnoses, prescription drugs, and provider orders. HAPUs were defined by PSI-03: stages III, IV, or unstageable pressure ulcers not present on admission as a secondary diagnosis, excluding cases of paralysis. Random forests reduced data dimensionality. Multilevel logistic regression of patient encounters evaluated associations between covariates and HAPU incidence. The approach produced a sample population of 21 153 patients with 1549 PSI-03 cases. The greatest odds ratio (OR) of HAPU incidence was among patients diagnosed with spinal cord injury (ICD-9 907.2: OR = 14.3; P  coded for paralysis, leading to a PSI-03 flag. Other high ORs included bed confinement (ICD-9 V49.84: OR = 3.1, P  coded without paralysis, leading to PSI-03 flags. The resulting statistical model can be tested to predict HAPUs during hospitalization. Inappropriate coding of conditions leads to poor hospital performance measures and Medicare reimbursement penalties. © 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

  16. A pulse coding and decoding strategy to perform Lamb wave inspections using simultaneously multiple actuators

    Science.gov (United States)

    De Marchi, Luca; Marzani, Alessandro; Moll, Jochen; Kudela, Paweł; Radzieński, Maciej; Ostachowicz, Wiesław

    2017-07-01

    The performance of Lamb wave based monitoring systems, both in terms of diagnosis time and data complexity, can be enhanced by increasing the number of transducers used to actuate simultaneously the guided waves in the inspected medium. However, in case of multiple simultaneously-operated actuators the interference among the excited wave modes within the acquired signals has to be considered for the further processing. To this aim, in this work a code division strategy based on the Warped Frequency Transform is presented. At first, the proposed procedure encodes actuation pulses using Gold sequences. Next, for each considered actuator the acquired signals are compensated from dispersion by cross correlating the warped version of the actuated and received signals. Compensated signals form the base for a final wavenumber imaging meant at emphasizing defects and or anomalies by removing incident wavefield and edge reflections. The proposed strategy is tested numerically, and validated through an experiment in which guided waves are actuated in a plate by four piezoelectric transducers operating simultaneously.

  17. Performance Test of CCTV in a Test Field

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Hyung Min [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2011-10-15

    On April 12-13, 2010, US President Obama hosted a Nuclear Security Summit in Washington, DC, to enhance international cooperation to prevent nuclear terrorism, an issue which he has identified as the most immediate and extreme threat to global security. The Summit focused on the security of nuclear materials, nonproliferation, disarmament, and peaceful nuclear energy. At the summit, the Republic of Korea was chosen as the host of the next Summit in 2012. This series of events reflects the growing global interest on 'Nuclear Security' and as the host country of the next Nuclear Summit it is the time for Korea to strengthen the physical protection regime for nuclear facilities as a first step of securing its nuclear security capability. KINAC has been operating Test field as a mean of preparing solid backup data for reviewing and revising DBT (Design Basis Threat) and to test components of the conventional physical protection system. CCTV is a key component which is used worldwide for the assessment measure of alarms. In terms of performance test of CCTV, there are several elements such as image quality, coverage and mechanical features (speed of zoom-in-out, capture, angle shift etc.). Speaking of image quality acquired by the CCTV, the quality is subject to resolution, monitor specification, camera housing, camera mounting and lightening. Thus it is clear that performance tests on image quality should consider those factors and vary the factors respectively in order to verify the influence and the interaction among those. Nevertheless due to the restrictions of the current Test field, this paper focuses on the image quality through resolution test under the various lightening conditions

  18. Eurados trial performance test for photon dosimetry

    DEFF Research Database (Denmark)

    Stadtmann, H.; Bordy, J.M.; Ambrosi, P.

    2001-01-01

    Within the framework of the EURADOS Action entitled Harmonisation and Dosimetric Quality Assurance in Individual Monitoring for External Radiation, trial performance tests for whole-body and extremity personal dosemeters were carried out. Photon, beta and neutron dosemeters were considered....... This paper summarises the results of the whole-body photon dosemeter test. Twenty-six dosimetry services from all EU Member States and Switzerland participated. Twelve different radiation fields were used to simulate various workplace irradiation fields. Dose values from 0.4 mSv to 80 mSv were chosen. From...

  19. FEMAXI-III, a computer code for fuel rod performance analysis

    International Nuclear Information System (INIS)

    Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

    1983-01-01

    This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

  20. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    Directory of Open Access Journals (Sweden)

    Giovedi Claudia

    2016-01-01

    Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

  1. Testing Solutions for Adult Film Performers.

    Science.gov (United States)

    Bergman, Zachary R

    2014-01-01

    The majority of the nation's adult films are produced in California, and within California, most production occurs in Los Angeles. In order to regulate that content, the County of Los Angeles passed the Safer Sex in the Adult Film Industry Act (Measure B) by way of referendum in November 2012. Measure B requires that adult film producers wishing to film in Los Angeles County obtain permits from the Los Angeles County Department of Public Health, and it also mandates that adult film performers use condoms while filming and "engaging in anal or vaginal sexual intercourse." Nevertheless, between August 2013 and January 2014, several adult film performers in California tested positive for HIV, and the threat of infection remains. Although Measure B is not the best way forward for Los Angeles County, elements of the ordinance should be incorporated into future legislative efforts. Given the economic ramifications of industry flight due to more localized regulations, this Note concludes that California should pass statewide comprehensive reform. Any such new legislation must treat "independent contractors," the classification generally used for adult film performs, as if they were regular employees. Legislation should also couple mandatory testing mechanisms with provisions granting performers the right to choose whether they use condoms. Finally, legislation must include mechanisms that ensure performers' preferences are not improperly tainted by outside forces and pressures. While there will always be risks associated with the production of adult content, if undertaken, these reforms could significantly mitigate those hazards.

  2. Performance test of 100 W linear compressor

    Energy Technology Data Exchange (ETDEWEB)

    Ko, J; Ko, D. Y.; Park, S. J.; Kim, H. B.; Hong, Y. J.; Yeom, H. K. [Korea Institute of Machinery and Materials, Daejeon(Korea, Republic of)

    2013-09-15

    In this paper, we present test results of developed 100 W class linear compressor for Stirling-type pulse tube refrigerator. The fabricated linear compressor has dual-opposed configuration, free piston and moving magnet type linear motor. Power transfer, efficiency and required pressure waveform are predicted with designed and measured specifications. In experiments, room temperature test with flow impedance is conducted to evaluate performance of developed linear compressor. Flow impedance is loaded to compressor with metering valve for flow resistance, inertance tube for flow inertance and buffer volumes for flow compliance. Several operating parameters such as input voltage, current, piston displacement and pressure wave are measured for various operating frequency and fixed input current level. Behaviors of dynamics and performance of linear compressor as varying flow impedance are discussed with measured experimental results. The developed linear compressor shows 124 W of input power, 86 % of motor efficiency and 60 % of compressor efficiency at its resonant operating condition.

  3. A ''SuperCode'' for performing systems analysis of tokamak experiments and reactors

    International Nuclear Information System (INIS)

    Haney, S.W.; Barr, W.L.; Crotinger, J.A.; Perkins, L.J.; Solomon, C.J.; Chaniotakis, E.A.; Freidberg, J.P.; Wei, J.; Galambos, J.D.; Mandrekas, J.

    1992-01-01

    A new code, named the ''SUPERCODE,'' has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of the physics and engineering models. The latter contain state-of-the-art plasma physics modelling but are limited in engineering content and time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2 dimensional MHD-transport plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formation, there is only a modest increase in CPU time over O-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, considerable effort has been expended to make the code user- and programming-friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community

  4. High performance optical encryption based on computational ghost imaging with QR code and compressive sensing technique

    Science.gov (United States)

    Zhao, Shengmei; Wang, Le; Liang, Wenqiang; Cheng, Weiwen; Gong, Longyan

    2015-10-01

    In this paper, we propose a high performance optical encryption (OE) scheme based on computational ghost imaging (GI) with QR code and compressive sensing (CS) technique, named QR-CGI-OE scheme. N random phase screens, generated by Alice, is a secret key and be shared with its authorized user, Bob. The information is first encoded by Alice with QR code, and the QR-coded image is then encrypted with the aid of computational ghost imaging optical system. Here, measurement results from the GI optical system's bucket detector are the encrypted information and be transmitted to Bob. With the key, Bob decrypts the encrypted information to obtain the QR-coded image with GI and CS techniques, and further recovers the information by QR decoding. The experimental and numerical simulated results show that the authorized users can recover completely the original image, whereas the eavesdroppers can not acquire any information about the image even the eavesdropping ratio (ER) is up to 60% at the given measurement times. For the proposed scheme, the number of bits sent from Alice to Bob are reduced considerably and the robustness is enhanced significantly. Meantime, the measurement times in GI system is reduced and the quality of the reconstructed QR-coded image is improved.

  5. The Multidimensional Influence of Acculturation on Digit Symbol-Coding and Wisconsin Card Sorting Test in Hispanics.

    Science.gov (United States)

    Krch, Denise; Lequerica, Anthony; Arango-Lasprilla, Juan Carlos; Rogers, Heather L; DeLuca, John; Chiaravalloti, Nancy D

    2015-01-01

    The purpose of the current study was to evaluate the relative contribution of acculturation to two tests of nonverbal test performance in Hispanics. This study compared 40 Hispanic and 20 non-Hispanic whites on Digit Symbol-Coding (DSC) and the Wisconsin Card Sorting Test (WCST) and evaluated the relative contribution of the various acculturation components to cognitive test performance in the Hispanic group. Hispanics performed significantly worse on DSC and WCST relative to non-Hispanic whites. Multiple regressions conducted within the Hispanic group revealed that language use uniquely accounted for 11.0% of the variance on the DSC, 18.8% of the variance on WCST categories completed, and 13.0% of the variance in perseverative errors on the WCST. Additionally, years of education in the United States uniquely accounted for 14.9% of the variance in DSC. The significant impact of acculturation on DSC and WCST lends support that nonverbal cognitive tests are not necessarily culture free. The differential contribution of acculturation proxies highlights the importance of considering these separate components when interpreting performance on neuropsychological tests in clinical and research settings. Factors, such as the country where education was received, may in fact be more meaningful information than the years of education of education attained. Thus, acculturation should be considered an important factor in any cognitive evaluation of culturally diverse individuals.

  6. Performance test of a TMS calorimeter

    International Nuclear Information System (INIS)

    Wild, B.

    1986-10-01

    Performance tests of a first calorimeter module using the room temperature liquid tetramethylsilane (TMS) as active element are described in detail. As absorber planed carbon steel slabs had been used. The charge yield is 70% of that in a very pure sample of the liquid. A long term stability of the signal with a lifetime of half a year has been realized. Experiences are described and the results explained in detail. (orig.) [de

  7. RHIC Sextant Test -- Physics and performance

    International Nuclear Information System (INIS)

    Wei, J.; Fischer, W.; Ahrens, L.

    1997-01-01

    This paper presents beam physics and machine performance results of the Relativistic Heavy Ion Collider (RHIC) Sextant and AGS-to-RHIC (AtR) transfer line during the Sextant Test in early 1997. Techniques used to measure both machine properties (difference orbits, dispersion, and beamline optics) and beam parameters (energy, intensity, transverse and longitudinal emittances) are described. Good agreement was achieved between measured and design lattice optics. The gold ion beam quality was shown to approach RHIC design requirements

  8. Performance tests on the NRPB thermoluminescent dosemeter

    CERN Document Server

    Shaw, K B

    1977-01-01

    Performance tests on the thermoluminescent dosemeter, designed at NRPB for use in the automated personal dosimetry system, are described. An ultra-thin lithium borate dosemeter has been developed for skin absorbed dose measurement. The X-ray, gamma-ray and beta-ray energy response of the dosemeter has been investigated and the angular response for the dosemeter has been examined. The annealing, read-out and stabilisation procedures for the dosemeter are described.

  9. The grout/glass performance assessment code system (GPACS) with verification and benchmarking

    International Nuclear Information System (INIS)

    Piepho, M.G.; Sutherland, W.H.; Rittmann, P.D.

    1994-12-01

    GPACS is a computer code system for calculating water flow (unsaturated or saturated), solute transport, and human doses due to the slow release of contaminants from a waste form (in particular grout or glass) through an engineered system and through a vadose zone to an aquifer, well and river. This dual-purpose document is intended to serve as a user's guide and verification/benchmark document for the Grout/Glass Performance Assessment Code system (GPACS). GPACS can be used for low-level-waste (LLW) Glass Performance Assessment and many other applications including other low-level-waste performance assessments and risk assessments. Based on all the cses presented, GPACS is adequate (verified) for calculating water flow and contaminant transport in unsaturated-zone sediments and for calculating human doses via the groundwater pathway

  10. 3D Analysis of Cooling Performance with Loss of Offsite Power Using GOTHIC Code

    International Nuclear Information System (INIS)

    Oh, Kye Min; Heo, Gyun Young; Na, In Sik; Choi, Yu Jung

    2010-01-01

    GOTHIC code enables to analyze one-dimensional or multi-dimensional problems for evaluating the cooling performance of loss of offsite power. The conventional GOTHIC code analysis performs heat transfer between plant containment and the outside of the fan cooler tubes by modeling each of fan cooler part model and component cooling water inside tube each to analyze boiling probability. In this paper, we suggest a way which reduces the multi-procedure of the cooling performance with loss of offsite power or the heat transfer states with complex geometrical structure to a single-procedure and verify the applicability of the heat transfer differences from the containment atmosphere humidity changes by the multi-nodes which component cooling water of tube or air of Reactor Containment Fan Cooler in the containment, otherwise the component model uses only one node

  11. Three computer codes to read, plot, and tabulate operational test-site recorded solar data. [TAPFIL, CHPLOT, and WRTCNL codes

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, S.D.; Sampson, R.J. Jr.; Stonemetz, R.E.; Rouse, S.L.

    1980-07-01

    A computer program, TAPFIL, has been developed by MSFC to read data from an IBM 360 tape for use on the PDP 11/70. The information (insolation, flowrates, temperatures, etc.) from 48 operational solar heating and cooling test sites is stored on the tapes. Two other programs, CHPLOT and WRTCNL, have been developed to plot and tabulate the data. These data will be used in the evaluation of collector efficiency and solar system performance. This report describes the methodology of the programs, their inputs, and their outputs.

  12. Simulation of international standard problem no. 44 open tests using Melcor computer code

    International Nuclear Information System (INIS)

    Song, Y.M.; Cho, S.W.

    2001-01-01

    MELCOR 1.8.4 code has been employed to simulate the KAEVER test series of K123/K148/K186/K188 that were proposed as open experiments of International Standard Problem No.44 by OECD-CSNI. The main purpose of this study is to evaluate the accuracy of the MELCOR aerosol model which calculates the aerosol distribution and settlement in a containment. For this, thermal hydraulic conditions are simulated first for the whole test period and then the behavior of hygroscopic CsOH/CsI and unsoluble Ag aerosols, which are predominant activity carriers in a release into the containment, is compared between the experimental results and the code predictions. The calculation results of vessel atmospheric concentration show a good simulation for dry aerosol but show large difference for wet aerosol due to a data mismatch in vessel humidity and the hygroscopicity. (authors)

  13. SASSYS-1 computer code verification with EBR-II test data

    International Nuclear Information System (INIS)

    Warinner, D.K.; Dunn, F.E.

    1985-01-01

    The EBR-II natural circulation experiment, XX08 Test 8A, is simulated with the SASSYS-1 computer code and the results for the latter are compared with published data taken during the transient at selected points in the core. The SASSYS-1 results provide transient temperature and flow responses for all points of interest simultaneously during one run, once such basic parameters as pipe sizes, initial core flows, and elevations are specified. The SASSYS-1 simulation results for the EBR-II experiment XX08 Test 8A, conducted in March 1979, are within the published plant data uncertainties and, thereby, serve as a partial verification/validation of the SASSYS-1 code

  14. Post test analysis of TEPSS tests -P2-, -P3-, -P5- and -P7- using the system code RELAP5/MOD 3.2

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.

    2000-01-01

    For the PANDA-Test-Facility (TEPSS configuration) post-test calculations and analyses have been performed for experiment -P2- (Early Start), -P3- (PCC start up), -P5- (Symmetric case, Two PCCs only) and -P7- (Severe Accident). Post test calculations have been performed with the system code RELAP5/Mod 3.2 using two different nodalization of the PANDA facility namely a basis nodalization and a much reduced one. The general trend of the calculations can be summarised: RELAP5/Mod3.2 calculated the general trends of the experiments sufficiently accurate; Using the reduced nodalization the results seem to be slightly more accurate than for the basic nodalization; On the other hand, calculations based on the reduced nodalization are not significantly faster than those with basic nodalization; The mass error is in the order of 200 to 900 kg. (author)

  15. Development and application of the BISON fuel performance code to the analysis of fission gas behaviour

    International Nuclear Information System (INIS)

    Pastore, G.; Hales, J.D.; Novascone, S.R.; Perez, D.M.; Spencer, B.W.; Williamson, R.L.

    2014-01-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that has been under development at Idaho National Laboratory (USA) since 2009. The capabilities of BISON comprise implicit solution of the fully coupled thermo-mechanics and diffusion equations, applicability to a variety of fuel forms, and simulation of both steady-state and transient conditions. The code includes multiphysics constitutive behavior for both fuel and cladding materials, and is designed for efficient use on highly parallel computers. This paper describes the main features of BISON, with emphasis on recent developments in modelling of fission gas behaviour in LWR-UO 2 fuel. The code is applied to the simulation of fuel rod irradiation experiments from the OECD/NEA International Fuel Performance Experiments Database. The comparison of the results with the available experimental data of fuel temperature, fission gas release, and cladding diametrical strain during pellet-cladding mechanical interaction is presented, pointing out a promising potential of the BISON code with the new fission gas behaviour model. (authors)

  16. 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.

  17. Performance and Complexity Evaluation of Iterative Receiver for Coded MIMO-OFDM Systems

    Directory of Open Access Journals (Sweden)

    Rida El Chall

    2016-01-01

    Full Text Available Multiple-input multiple-output (MIMO technology in combination with channel coding technique is a promising solution for reliable high data rate transmission in future wireless communication systems. However, these technologies pose significant challenges for the design of an iterative receiver. In this paper, an efficient receiver combining soft-input soft-output (SISO detection based on low-complexity K-Best (LC-K-Best decoder with various forward error correction codes, namely, LTE turbo decoder and LDPC decoder, is investigated. We first investigate the convergence behaviors of the iterative MIMO receivers to determine the required inner and outer iterations. Consequently, the performance of LC-K-Best based receiver is evaluated in various LTE channel environments and compared with other MIMO detection schemes. Moreover, the computational complexity of the iterative receiver with different channel coding techniques is evaluated and compared with different modulation orders and coding rates. Simulation results show that LC-K-Best based receiver achieves satisfactory performance-complexity trade-offs.

  18. Calculations of Edwards' pipe blowdown tests using the code TRAC P1

    International Nuclear Information System (INIS)

    O'Mahoney, R.

    1979-05-01

    The paper describes the results obtained using the non-thermal equilibrium LOCA code TRAC-P1 for two of a series of Pipe Blowdown Tests. Comparisons are made with the experimental values and RELAP-UK Mark IV predictions. Some discrepancies between prediction and experiment are observed, and certain aspects of the model are considered to warrant possible further attention. (U.K.)

  19. Coded excitation for infrared non-destructive testing of carbon fiber reinforced plastics.

    Science.gov (United States)

    Mulaveesala, Ravibabu; Venkata Ghali, Subbarao

    2011-05-01

    This paper proposes a Barker coded excitation for defect detection using infrared non-destructive testing. Capability of the proposed excitation scheme is highlighted with recently introduced correlation based post processing approach and compared with the existing phase based analysis by taking the signal to noise ratio into consideration. Applicability of the proposed scheme has been experimentally validated on a carbon fiber reinforced plastic specimen containing flat bottom holes located at different depths.

  20. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  1. TEMPEST code modifications and testing for erosion-resisting sludge simulations

    International Nuclear Information System (INIS)

    Onishi, Y.; Trent, D.S.

    1998-01-01

    The TEMPEST computer code has been used to address many waste retrieval operational and safety questions regarding waste mobilization, mixing, and gas retention. Because the amount of sludge retrieved from the tank is directly related to the sludge yield strength and the shear stress acting upon it, it is important to incorporate the sludge yield strength into simulations of erosion-resisting tank waste retrieval operations. This report describes current efforts to modify the TEMPEST code to simulate pump jet mixing of erosion-resisting tank wastes and the models used to test for erosion of waste sludge with yield strength. Test results for solid deposition and diluent/slurry jet injection into sludge layers in simplified tank conditions show that the modified TEMPEST code has a basic ability to simulate both the mobility and immobility of the sludges with yield strength. Further testing, modification, calibration, and verification of the sludge mobilization/immobilization model are planned using erosion data as they apply to waste tank sludges

  2. Performance Measures of Diagnostic Codes for Detecting Opioid Overdose in the Emergency Department.

    Science.gov (United States)

    Rowe, Christopher; Vittinghoff, Eric; Santos, Glenn-Milo; Behar, Emily; Turner, Caitlin; Coffin, Phillip O

    2017-04-01

    Opioid overdose mortality has tripled in the United States since 2000 and opioids are responsible for more than half of all drug overdose deaths, which reached an all-time high in 2014. Opioid overdoses resulting in death, however, represent only a small fraction of all opioid overdose events and efforts to improve surveillance of this public health problem should include tracking nonfatal overdose events. International Classification of Disease (ICD) diagnosis codes, increasingly used for the surveillance of nonfatal drug overdose events, have not been rigorously assessed for validity in capturing overdose events. The present study aimed to validate the use of ICD, 9th revision, Clinical Modification (ICD-9-CM) codes in identifying opioid overdose events in the emergency department (ED) by examining multiple performance measures, including sensitivity and specificity. Data on ED visits from January 1, 2012, to December 31, 2014, including clinical determination of whether the visit constituted an opioid overdose event, were abstracted from electronic medical records for patients prescribed long-term opioids for pain from any of six safety net primary care clinics in San Francisco, California. Combinations of ICD-9-CM codes were validated in the detection of overdose events as determined by medical chart review. Both sensitivity and specificity of different combinations of ICD-9-CM codes were calculated. Unadjusted logistic regression models with robust standard errors and accounting for clustering by patient were used to explore whether overdose ED visits with certain characteristics were more or less likely to be assigned an opioid poisoning ICD-9-CM code by the documenting physician. Forty-four (1.4%) of 3,203 ED visits among 804 patients were determined to be opioid overdose events. Opioid-poisoning ICD-9-CM codes (E850.2-E850.2, 965.00-965.09) identified overdose ED visits with a sensitivity of 25.0% (95% confidence interval [CI] = 13.6% to 37.8%) and

  3. Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

    International Nuclear Information System (INIS)

    Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

    1993-09-01

    PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

  4. MELMRK 2.0: A description of computer models and results of code testing

    International Nuclear Information System (INIS)

    Wittman, R.S.; Denny, V.; Mertol, A.

    1992-01-01

    An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly

  5. Test Program for the Performance Analysis of DNS64 Servers

    Directory of Open Access Journals (Sweden)

    Gábor Lencse

    2015-09-01

    Full Text Available In our earlier research papers, bash shell scripts using the host Linux command were applied for testing the performance and stability of different DNS64 server imple­mentations. Because of their inefficiency, a small multi-threaded C/C++ program (named dns64perf was written which can directly send DNS AAAA record queries. After the introduction to the essential theoretical background about the structure of DNS messages and TCP/IP socket interface programming, the design decisions and implementation details of our DNS64 performance test program are disclosed. The efficiency of dns64perf is compared to that of the old method using bash shell scripts. The result is convincing: dns64perf can send at least 95 times more DNS AAAA record queries per second. The source code of dns64perf is published under the GNU GPLv3 license to support the work of other researchers in the field of testing the performance of DNS64 servers.

  6. Performance Tests for Bubble Blockage Device

    International Nuclear Information System (INIS)

    Ha, Kwang Soon; Wi, Kyung Jin; Park, Rae Joon; Wan, Han Seong

    2014-01-01

    Postulated severe core damage accidents have a high threat risk for the safety of human health and jeopardize the environment. Versatile measures have been suggested and applied to mitigate severe accidents in nuclear power plants. To improve the thermal margin for the severe accident measures in high-power reactors, engineered corium cooling systems involving boiling-induced two-phase natural circulation have been proposed for decay heat removal. A boiling-induced natural circulation flow is generated in a coolant path between a hot vessel wall and cold coolant reservoir. In general, it is possible for some bubbles to be entrained in the natural circulation loop. If some bubbles entrain in the liquid phase flow passage, flow instability may occur, that is, the natural circulation mass flow rate may be oscillated. A new device to block the entraining bubbles is proposed and verified using air-water test loop. To avoid bubbles entrained in the natural circulation flow loop, a new device was proposed and verified using an air-water test loop. The air injection and liquid circulation loop was prepared, and the tests for the bubble blockage devices were performed by varying the geometry and shape of the devices. The performance of the bubble blockage device was more effective as the area ratio of the inlet to the down-comer increased, and the device height decreased. If the device has a rim to generate a vortex zone, the bubbles will be most effectively blocked

  7. Performance testing SPECT systems: Recent developments

    International Nuclear Information System (INIS)

    Todd-Pokropek, A.

    1985-01-01

    A protocol is used to test as many currently available systems as possible, including both rotating gamma camera systems from GE, Siemens, Toshiba, Technicare, Elscint, Philips, Picker, etc, and single/multi slice system such as the Medimatic. This paper indicates changes that have occurred with respect to preliminary results previously published. Reconstructions were performed in a standard manner, using a Ramp filter and without attenuation correction, in order to reduce the effect of differences in software. The Jaszczak phantom was used to assess overall image quality. Tests were performed using the different collimators available. Image quality improved with better spatial resolution, even at the expense of sensitivity. Non-circular orbits, where available, were tested, and resolution was comparing for various positions and directions. With double headed systems, the results obtained were compared to those using a single head alone. Some gain in image quality seems to have been achieved by better uniformity resulting from improved correction circuitry, by the use of non-circular orbits, and from scatter correction. Potential still exists to improve collimator design

  8. Ethics Standards Impacting Test Development and Use: A Review of 31 Ethics Codes Impacting Practices in 35 Countries

    Science.gov (United States)

    Leach, Mark M.; Oakland, Thomas

    2007-01-01

    Ethics codes are designed to protect the public by prescribing behaviors professionals are expected to exhibit. Although test use is universal, albeit reflecting strong Western influences, previous studies that examine the degree issues pertaining to test development and use and that are addressed in ethics codes of national psychological…

  9. Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

    International Nuclear Information System (INIS)

    Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

    2006-01-01

    as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined. (authors)

  10. DPM PERFORMANCE TESTING USING RASPBERRY PIS

    CERN Document Server

    Regala, M

    2013-01-01

    This is the final report from attending CERN’s Summer Student Programme. The project goal was to do performance testing on the Disk Pool Manager (DPM), a lightweight, reliable, grid-aware storage software used to store and retrieve data produced by CERN’s LHC experiments using the small, low-end ARM powered devices named Raspberry Pis. The idea behind it was to reason if it’s possible to use a cluster of lower-end, under-capable devices to run DPM, and to conclude if it would be more energy efficient than running it on oversized machines, with the same or comparable performance. If this hypothesis was true, the power-hungry machines could be ditched in favour of these small devices, leading to an enormous saving in overall power consumption and hence, overall cost. In this report, I describe what was the initial project goal and intended outcomes, proceeding to explain the underlying technologies used. Afterwards, I’ll explain the setup used, the tests performed, and the conclusions reached. iii

  11. Interpretation, with respect to ASME code Case N-318, of limit moment and fatigue tests of lugs welded to pipe

    International Nuclear Information System (INIS)

    Foster, D.C.; Van Duyne, D.A.; Budlong, L.A.; Muffett, J.W.; Wais, E.A.; Streck, G.; Rodabaugh, E.C.

    1990-01-01

    Two nonmandatory ASME code cases have been used often in the evaluation of lugs on nuclear-power- plant piping systems. ASME Code Case N-318 provides guidance for evaluation of the design of rectangular cross-section attachments on Class 2 or 3 piping, and ASME Code Case N-122 provides guidance for evaluation of lugs on Class 1 piping. These code cases have been reviewed and evaluated based on available test data. The results indicate that the Code cases are overly conservative. Recommendations for revisions to the cases are presented which, if adopted, will reduce the overconservatism

  12. Equation-of-State Test Suite for the DYNA3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, Russell D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-11-05

    This document describes the creation and implementation of a test suite for the Equationof- State models in the DYNA3D code. A customized input deck has been created for each model, as well as a script that extracts the relevant data from the high-speed edit file created by DYNA3D. Each equation-of-state model is broken apart and individual elements of the model are tested, as well as testing the entire model. The input deck for each model is described and the results of the tests are discussed. The intent of this work is to add this test suite to the validation suite presently used for DYNA3D.

  13. SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

    International Nuclear Information System (INIS)

    Baker, R.B.; Wilson, D.R.

    1986-04-01

    The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

  14. Validation of ASTEC v1.0 computer code against FPT2 test

    International Nuclear Information System (INIS)

    Mladenov, I.; Tusheva, P.; Kalchev, B.; Dimov, D.; Ivanov, I.

    2005-01-01

    The aim of the work is by various nodalization schemes of the model to investigate the ASTEC v1.0 computer code sensitivity and to validate the code against PHEBUS - FPT2 experiment. This code is used for severe accident analysis. The aim corresponds to the main technical objective of the experiment which is to contribute to the validation of models and computer codes to be used for the calculation of the source term in case of a severe accident in a Light Water Reactor. The objective's scope of the FPT2 is large - separately for the bundle, the experimental circuit and the containment. Additional objectives are to characterize aerosol sizing and deposition processes, and also potential FP poisoning effects on hydrogen recombiner coupons exposed to containment atmospheric conditions representative of a LWR severe accident. The analyses of the results of the performed calculations show a good accordance with the reference case calculations, and then with the experimental data. Some differences in the calculations for the thermal behavior appear locally during the oxidation phase and the heat-up phase. There is very good confirmation regarding the volatile and semi-volatile fission products release from the fuel pellets. Important for analysis of the process is the final axial distribution of the mass of fuel relocation obtained at the end of the calculation

  15. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  16. 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.

  17. Automated JPSS VIIRS GEO code change testing by using Chain Run Scripts

    Science.gov (United States)

    Chen, W.; Wang, W.; Zhao, Q.; Das, B.; Mikles, V. J.; Sprietzer, K.; Tsidulko, M.; Zhao, Y.; Dharmawardane, V.; Wolf, W.

    2015-12-01

    The Joint Polar Satellite System (JPSS) is the next generation polar-orbiting operational environmental satellite system. The first satellite in the JPSS series of satellites, J-1, is scheduled to launch in early 2017. J1 will carry similar versions of the instruments that are on board of Suomi National Polar-Orbiting Partnership (S-NPP) satellite which was launched on October 28, 2011. The center for Satellite Applications and Research Algorithm Integration Team (STAR AIT) uses the Algorithm Development Library (ADL) to run S-NPP and pre-J1 algorithms in a development and test mode. The ADL is an offline test system developed by Raytheon to mimic the operational system while enabling a development environment for plug and play algorithms. The Perl Chain Run Scripts have been developed by STAR AIT to automate the staging and processing of multiple JPSS Sensor Data Record (SDR) and Environmental Data Record (EDR) products. JPSS J1 VIIRS Day Night Band (DNB) has anomalous non-linear response at high scan angles based on prelaunch testing. The flight project has proposed multiple mitigation options through onboard aggregation, and the Option 21 has been suggested by the VIIRS SDR team as the baseline aggregation mode. VIIRS GEOlocation (GEO) code analysis results show that J1 DNB GEO product cannot be generated correctly without the software update. The modified code will support both Op21, Op21/26 and is backward compatible with SNPP. J1 GEO code change version 0 delivery package is under development for the current change request. In this presentation, we will discuss how to use the Chain Run Script to verify the code change and Lookup Tables (LUTs) update in ADL Block2.

  18. Parameters that affect parallel processing for computational electromagnetic simulation codes on high performance computing clusters

    Science.gov (United States)

    Moon, Hongsik

    What is the impact of multicore and associated advanced technologies on computational software for science? Most researchers and students have multicore laptops or desktops for their research and they need computing power to run computational software packages. Computing power was initially derived from Central Processing Unit (CPU) clock speed. That changed when increases in clock speed became constrained by power requirements. Chip manufacturers turned to multicore CPU architectures and associated technological advancements to create the CPUs for the future. Most software applications benefited by the increased computing power the same way that increases in clock speed helped applications run faster. However, for Computational ElectroMagnetics (CEM) software developers, this change was not an obvious benefit - it appeared to be a detriment. Developers were challenged to find a way to correctly utilize the advancements in hardware so that their codes could benefit. The solution was parallelization and this dissertation details the investigation to address these challenges. Prior to multicore CPUs, advanced computer technologies were compared with the performance using benchmark software and the metric was FLoting-point Operations Per Seconds (FLOPS) which indicates system performance for scientific applications that make heavy use of floating-point calculations. Is FLOPS an effective metric for parallelized CEM simulation tools on new multicore system? Parallel CEM software needs to be benchmarked not only by FLOPS but also by the performance of other parameters related to type and utilization of the hardware, such as CPU, Random Access Memory (RAM), hard disk, network, etc. The codes need to be optimized for more than just FLOPs and new parameters must be included in benchmarking. In this dissertation, the parallel CEM software named High Order Basis Based Integral Equation Solver (HOBBIES) is introduced. This code was developed to address the needs of the

  19. Performance testing of a hydrogen heat pipe

    International Nuclear Information System (INIS)

    Alario, J.; Kosson, R.

    1980-01-01

    Test results are presented for a reentrant groove heat pipe with hydrogen working fluid. The heat pipe became operational between 20 and 30 K after a cooldown from 77 K without any difficulty. Steady-state performance data taken over a 19 to 23 K temperature range indicated the following: (1) maximum heat transport capacity 5.4 W-m (2) static wicking height 1.42 cm and (3) overall heat pipe conductance 1.7 W/C. These data agreed remarkably well with extrapolations made from comparable ammonia test results. The maximum heat transport capacity is 9.5% larger than the extrapolated value, but the static wicking height is the same. The overall conductance is 29% of the ammonia value, which is close to the ratio of liquid thermal conductivities (24%). Also, recovery from a completely frozen condition was accomplished within 5 min by simply applying an evaporater heat load of 1.8 W

  20. Compilation of Quality Assurance Documentation for Analyses Performed for the Resumption of Transient Testing Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Schafer, Annette L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sondrup, A. Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-11-01

    This is a companion document to the analyses performed in support of the environmental assessment for the Resumption of Transient Fuels and Materials Testing. It is provided to allow transparency of the supporting calculations. It provides computer code input and output. The basis for the calculations is documented separately in INL (2013) and is referenced, as appropriate. Spreadsheets used to manipulate the code output are not provided.

  1. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  2. Radon detection system, design, test and performance

    International Nuclear Information System (INIS)

    Balcazar, M.; Chavez, A.; Pina-Villalpando, G.; Navarrete, M.

    1999-01-01

    A portable radon detection system (α-Inin) has been designed and constructed for using it in adverse environmental conditions where humidity, temperature and chemical vaporous are present. The minimum integration time is in periods of 15 min during 41 days. A 12 V battery and a photovoltaic module allow the α-Inin autonomy in field measurements. Data is collected by means of a laptop computer where data processing and α-Inin programming are carried out. α-Inin performance was simultaneously tested in a controlled radon chamber, together with a commercial α-Meter

  3. Performance test of wet type decontamination device

    International Nuclear Information System (INIS)

    Lee, E. P.; Kim, E. G.; Min, D. K.; Jun, Y. B.; Lee, H. K.; Seu, H. S.; Kwon, H. M.; Hong, K.P.

    2003-01-01

    The intervention area located at rear hot cell can be contaminated by hot cell maintenance work. For effective decontamination of the intervention floor a wet type decontamination device was developed. The device was assembled with a brush rotating part, a washing liquid supplying part, an intake part for recovering contaminated liquid and a device moving cart part. The device was made of stainless steel for easy decontamination and corrosion resistance. The function test carried out at intervention area of the PIE facility showed good performance

  4. Performance testing of UK personal dosimetry laboratories

    CERN Document Server

    Marshall, T O

    1985-01-01

    The proposed Ionising Radiations Regulations will require all UK personal dosimetry laboratories that monitor classified personnel to be approved for personal dosimetry by the Health and Safety Executive. It is suggested that these approvals should be based on general and supplementary criteria published by the British Calibration Service (BCS) for laboratory approval for the provision of personal dosimetry services. These criteria specify certain qualitative requirements and also indicate the need for regular tests of performance to be carried out to ensure constancy of dosimetric standards. This report concerns the latter. The status of the BCS criteria is discussed and the need for additional documents to cover new techniques and some modifications to existing documents is indicated. A means is described by which the technical performance of laboratories, concerned with personal monitoring for external radiations, can be assessed, both initially and ongoing. The costs to establish the scheme and operate it...

  5. Performance testing of UK personal dosimetry laboratories

    International Nuclear Information System (INIS)

    Marshall, T.O.

    1985-01-01

    The proposed Ionising Radiations Regulations will require all UK personal dosimetry laboratories that monitor classified personnel to be approved for personal dosimetry by the Health and Safety Executive. It is suggested that these approvals should be based on general and supplementary criteria published by the British Calibration Service (BCS) for laboratory approval for the provision of personal dosimetry services. These criteria specify certain qualitative requirements and also indicate the need for regular tests of performance to be carried out to ensure constancy of dosimetric standards. This report concerns the latter. The status of the BCS criteria is discussed and the need for additional documents to cover new techniques and some modifications to existing documents is indicated. A means is described by which the technical performance of laboratories, concerned with personal monitoring for external radiations, can be assessed, both initially and ongoing. The costs to establish the scheme and operate it are also estimated. (author)

  6. Post-test sensitivity analysis of OECD/CSNI ISP42 panda experiment by Relap5 code

    International Nuclear Information System (INIS)

    Zanocco, P.; D'Auria, F.; Galassi, G.M.

    2001-01-01

    The present document deals with Relap5/Mod3.2 analysis of the International Standard Problem (ISP-42) exercise performed in PANDA facility on April 21-22, 1998. PANDA is installed at PSI (Paul Scherrer Institute). PANDA is a large-scale thermal-hydraulic test facility suitable for the simulation of passive containment for Advanced Light Water Reactors (ALWR). The work focuses phase A of the ISP-42 experiment, including the break in the main steam line, and the Passive Containment Cooling System Start-Up. The objective is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air and to observe the resulting system behavior. A detailed nodalization was set-up at the University of Pisa, in order to model 3-D flow paths with a 1-D code. The comparison between pre-test predictions and experimental data is discussed. Overall time behavior is reasonably well predicted, showing a rather good and robust overall code behavior in the simulation of the global test scenario. The results of a preliminary post-test analysis are discussed, including the comparison with the experimental data. (authors)

  7. Model tests of a once-through steam generator for land-blocker assessment and THEDA code verification. Final report

    International Nuclear Information System (INIS)

    Carter, H.R.; Childerson, M.T.; Moskal, T.E.

    1983-06-01

    The Babcock and Wilcox Company (B and W) operating Once-Through Steam Generators (OTSGs) have experienced leaking tubes in a region adjacent to the untubed inspection lane. The tube leaks have been attributed to an environmentally-assisted fatigue mechanism with moisture transported up the inspection lane being a major factor in the tube-failure process. B and W has developed a hardware modification (lane blockers) to mitigate the detrimental effects of inspection lane moisture. A 30-tube Laboratory Once-through Steam Generator (Designated OTSGC) was designed, fabricated, and tested. Tests were performed with and without five flat-plate lane blockers installed on tube-support plates (TSPs) 10, 11, 12, 13, and 14. The test results were utilized to determine the effectiveness of lane blockers for eliminating moisture transport to the upper tubesheet in the inspection lanes and to benchmark the predictive capabilities of a three-dimensional steam-generator computer code, THEDA

  8. Performance Based Plastic Design of Concentrically Braced Frame attuned with Indian Standard code and its Seismic Performance Evaluation

    Directory of Open Access Journals (Sweden)

    Sejal Purvang Dalal

    2015-12-01

    Full Text Available In the Performance Based Plastic design method, the failure is predetermined; making it famous throughout the world. But due to lack of proper guidelines and simple stepwise methodology, it is not quite popular in India. In this paper, stepwise design procedure of Performance Based Plastic Design of Concentrically Braced frame attuned with the Indian Standard code has been presented. The comparative seismic performance evaluation of a six storey concentrically braced frame designed using the displacement based Performance Based Plastic Design (PBPD method and currently used force based Limit State Design (LSD method has also been carried out by nonlinear static pushover analysis and time history analysis under three different ground motions. Results show that Performance Based Plastic Design method is superior to the current design in terms of displacement and acceleration response. Also total collapse of the frame is prevented in the PBPD frame.

  9. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, Peter Andrew

    2011-12-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  10. PERFORMANCE EVALUATION OF TURBO CODED OFDM SYSTEMS AND APPLICATION OF TURBO DECODING FOR IMPULSIVE CHANNEL

    Directory of Open Access Journals (Sweden)

    Savitha H. M.

    2010-09-01

    Full Text Available A comparison of the performance of hard and soft-decision turbo coded Orthogonal Frequency Division Multiplexing systems with Quadrature Phase Shift Keying (QPSK and 16-Quadrature Amplitude Modulation (16-QAM is considered in the first section of this paper. The results show that the soft-decision method greatly outperforms the hard-decision method. The complexity of the demapper is reduced with the use of simplified algorithm for 16-QAM demapping. In the later part of the paper, we consider the transmission of data over additive white class A noise (AWAN channel, using turbo coded QPSK and 16-QAM systems. We propose a novel turbo decoding scheme for AWAN channel. Also we compare the performance of turbo coded systems with QPSK and 16-QAM on AWAN channel with two different channel values- one computed as per additive white Gaussian noise (AWGN channel conditions and the other as per AWAN channel conditions. The results show that the use of appropriate channel value in turbo decoding helps to combat the impulsive noise more effectively. The proposed model for AWAN channel exhibits comparable Bit error rate (BER performance as compared to AWGN channel.

  11. High-performance computational fluid dynamics: a custom-code approach

    International Nuclear Information System (INIS)

    Fannon, James; Náraigh, Lennon Ó; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain

    2016-01-01

    We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier–Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing. (paper)

  12. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC)

    International Nuclear Information System (INIS)

    Schultz, Peter Andrew

    2011-01-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  13. Source-term model for the SYVAC3-NSURE performance assessment code

    International Nuclear Information System (INIS)

    Rowat, J.H.; Rattan, D.S.; Dolinar, G.M.

    1996-11-01

    Radionuclide contaminants in wastes emplaced in disposal facilities will not remain in those facilities indefinitely. Engineered barriers will eventually degrade, allowing radioactivity to escape from the vault. The radionuclide release rate from a low-level radioactive waste (LLRW) disposal facility, the source term, is a key component in the performance assessment of the disposal system. This report describes the source-term model that has been implemented in Ver. 1.03 of the SYVAC3-NSURE (Systems Variability Analysis Code generation 3-Near Surface Repository) code. NSURE is a performance assessment code that evaluates the impact of near-surface disposal of LLRW through the groundwater pathway. The source-term model described here was developed for the Intrusion Resistant Underground Structure (IRUS) disposal facility, which is a vault that is to be located in the unsaturated overburden at AECL's Chalk River Laboratories. The processes included in the vault model are roof and waste package performance, and diffusion, advection and sorption of radionuclides in the vault backfill. The model presented here was developed for the IRUS vault; however, it is applicable to other near-surface disposal facilities. (author). 40 refs., 6 figs

  14. High-performance computational fluid dynamics: a custom-code approach

    Science.gov (United States)

    Fannon, James; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain; Náraigh, Lennon Ó.

    2016-07-01

    We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier-Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing.

  15. Performance testing of real-time AI systems using the activation framework

    International Nuclear Information System (INIS)

    Becker, L.; Duckworth, J.; Laznovsky, A.; Green, P.

    1992-01-01

    This paper describes methods for automated performance testing of real-time artificial intelligence systems using the Activation Framework software development tool. The Activation Framework is suitable for applications such as the diagnosis of power system failures, which require the interpretation of large volumes of data in real-time. The Activation Framework consists of tools for compiling groups of Expert Systems rules into executable code modules, for automatically generating code modules from high level system configuration descriptions, and for automatically generating command files for program compilation and linking. It includes an operating system environment which provides the code which is common from one real-time AI applications to the next. It also includes mechanisms, described here, for automatic performance testing. The principal emphasis of this paper is on a rule based language which is used to capture performance specifications. This specification is compiled into code modules which are used to automatically test the system. This testing can validate that the system meets performance requirements during development and after maintenance. A large number of tests can be randomly generated and executed and the correctness of the outputs automatically validated. The paper also describes graph directed testing methods to minimize the number of test runs required

  16. 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

  17. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Manolova, M.; Stefanova, S.; Simeonova, V.; Passage, G.; Lassmann, K.

    1994-01-01

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: 1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; 2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; 3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs

  18. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Vitkova, M; Manolova, M; Stefanova, S; Simeonova, V; Passage, G [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Lassmann, K [European Atomic Energy Community, Karlsruhe (Germany). European Inst. for Transuranium Elements

    1994-12-31

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: (1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; (2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; (3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs.

  19. Assessing the Predictive Capability of the LIFEIV Nuclear Fuel Performance Code using Sequential Calibration

    International Nuclear Information System (INIS)

    Stull, Christopher J.; Williams, Brian J.; Unal, Cetin

    2012-01-01

    This report considers the problem of calibrating a numerical model to data from an experimental campaign (or series of experimental tests). The issue is that when an experimental campaign is proposed, only the input parameters associated with each experiment are known (i.e. outputs are not known because the experiments have yet to be conducted). Faced with such a situation, it would be beneficial from the standpoint of resource management to carefully consider the sequence in which the experiments are conducted. In this way, the resources available for experimental tests may be allocated in a way that best 'informs' the calibration of the numerical model. To address this concern, the authors propose decomposing the input design space of the experimental campaign into its principal components. Subsequently, the utility (to be explained) of each experimental test to the principal components of the input design space is used to formulate the sequence in which the experimental tests will be used for model calibration purposes. The results reported herein build on those presented and discussed in (1,2) wherein Verification and Validation and Uncertainty Quantification (VU) capabilities were applied to the nuclear fuel performance code LIFEIV. In addition to the raw results from the sequential calibration studies derived from the above, a description of the data within the context of the Predictive Maturity Index (PMI) will also be provided. The PMI (3,4) is a metric initiated and developed at Los Alamos National Laboratory to quantitatively describe the ability of a numerical model to make predictions in the absence of experimental data, where it is noted that 'predictions in the absence of experimental data' is not synonymous with extrapolation. This simply reflects the fact that resources do not exist such that each and every execution of the numerical model can be compared against experimental data. If such resources existed, the justification for numerical models

  20. Influence of Code Size Variation on the Performance of 2D Hybrid ZCC/MD in OCDMA System

    Directory of Open Access Journals (Sweden)

    Matem Rima.

    2018-01-01

    Full Text Available Several two dimensional OCDMA have been developed in order to overcome many problems in optical network, enhancing cardinality, suppress Multiple Access Interference (MAI and mitigate Phase Induced Intensity Noise (PIIN. This paper propose a new 2D hybrid ZCC/MD code combining between 1D ZCC spectral encoding where M is its code length and 1D MD spatial spreading where N is its code length. The spatial spreading (N code length offers a good cardinality so it represents the main effect to enhance the performance of the system compared to the spectral (M code length according to the numerical results.

  1. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    Energy Technology Data Exchange (ETDEWEB)

    Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  2. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

    2017-01-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  3. Standard specification for agencies performing nondestructive testing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This specification covers minimum requirements for agencies performing nondestructive testing (NDT). 1.2 When using this specification to assess the capability of, or to accredit NDT agencies, Guide E 1359 shall be used as a basis for the survey. It can be supplemented as necessary with more detail in order to meet the auditor's specific needs. 1.3 This specification can be used as a basis to evaluate testing or inspection agencies, or both, and is intended for use for the qualifying or accrediting, or both, of testing or inspection agencies, public or private. 1.4 The use of SI or inch-pound units, or combination thereof, will be the responsibility of the technical committee whose standards are referred to in this standard. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to...

  4. Development of European creep crack growth testing code of practice for industrial specimens

    International Nuclear Information System (INIS)

    Dogan, B.; Nikbin, K.; Petrovski, B.

    2004-01-01

    The integrity and residual life assessment of high temperature components require defects, detected or assumed to exist, through minimum allowable limits of detectable flaws using nondestructive testing methods. It relies on information obtained from the material's mechanical, uniaxial creep, creep crack initiation and growth properties. The information derived from experiments needs to be validated and harmonised following a Code of Practice that data variability between different institutions can be reduced to a minimum. The present paper reports on a Code of Practice (CoP) being prepared within the framework of the partially European Commission funded project CRETE. The novel aspect of the presented CoP is the inclusion of component relevant industrial specimen geometries. It covers testing and analysis of Creep Crack growth (CCG) in metallic materials at elevated temperature using six different cracked geometries that have been validated in. It aims to give advice on testing, measurements and analysis of creep crack growth data for a range of creep brittle to creep ductile materials using component service relevant specimen geometries and sizes. The CoP may be used for material selection criteria and inspection requirements for damage tolerant applications. In quantitative terms, these types of tests can be used to assess the individual and combined effects of metallurgical, fabrication, operating temperature, and loading conditions on creep crack growth life. Further issues will be addressed including material properties, damage and crack growth related constraint effect, stress relaxation and stress-strain fields, residual stresses, partitioning displacement, analysis of elastic creep, elastic compliance measurements

  5. Fenestration System Performance Research, Testing, and Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Jim Benney

    2009-11-30

    The US DOE was and is instrumental to NFRC's beginning and its continued success. The 2005 to 2009 funding enables NFRC to continue expanding and create new, improved ratings procedures. Research funded by the US DOE enables increased fenestration energy rating accuracy. International harmonization efforts supported by the US DOE allow the US to be the global leader in fenestration energy ratings. Many other governments are working with the NFRC to share its experience and knowledge toward development of their own national fenestration rating process similar to the NFRC's. The broad and diverse membership composition of NFRC allows anyone with a fenestration interest to come forward with an idea or improvement to the entire fenestration community for consideration. The NFRC looks forward to the next several years of growth while remaining the nation's resource for fair, accurate, and credible fenestration product energy ratings. NFRC continues to improve its rating system by considering new research, methodologies, and expanding to include new fenestration products. Currently, NFRC is working towards attachment energy ratings. Attachments are blinds, shades, awnings, and overhangs. Attachments may enable a building to achieve significant energy savings. An NFRC rating will enable fair competition, a basis for code references, and a new ENERGY STAR product category. NFRC also is developing rating methods to consider non specular glazing such as fritted glass. Commercial applications frequently use fritted glazing, but no rating method exists. NFRC is testing new software that may enable this new rating and contribute further to energy conservation. Around the world, many nations are seeking new energy conservation methods and NFRC is poised to harmonize its rating system assisting these nations to better manage and conserve energy in buildings by using NFRC rated and labeled fenestration products. As this report has shown, much more work needs to be

  6. Application of the BISON Fuel Performance Code of the FUMEX-III Coordinated Research Project

    International Nuclear Information System (INIS)

    Williamson, R.L.; Novascone, S.R.

    2013-01-01

    Since 1981, the International Atomic Energy Agency (IAEA) has sponsored a series of Coordinated Research Projects (CRP) in the area of nuclear fuel modeling. These projects have typically lasted 3-5 years and have had broad international participation. The objectives of the projects have been to assess the maturity and predictive capability of fuel performance codes, support interaction and information exchange between countries with code development and application needs, build a database of well- defined experiments suitable for code validation, transfer a mature fuel modeling code to developing countries, and provide guidelines for code quality assurance and code application to fuel licensing. The fourth and latest of these projects, known as FUMEX-III1 (FUel Modeling at EXtended Burnup- III), began in 2008 and ended in December of 2011. FUMEX-III was the first of this series of fuel modeling CRP's in which the INL participated. Participants met at the beginning of the project to discuss and select a set of experiments ('priority cases') for consideration during the project. These priority cases were of broad interest to the participants and included reasonably well-documented and reliable data. A meeting was held midway through the project for participants to present and discuss progress on modeling the priority cases. A final meeting was held at close of the project to present and discuss final results and provide input for a final report. Also in 2008, the INL initiated development of a new multidimensional (2D and 3D) multiphysics nuclear fuel performance code called BISON, with code development progressing steadily during the three-year FUMEX-III project. Interactions with international fuel modeling researchers via FUMEX-III played a significant role in the BISON evolution, particularly influencing the selection of material and behavioral models which are now included in the code. The FUMEX-III cases are generally integral fuel rod experiments occurring

  7. Los Alamos and Lawrence Livermore National Laboratories Code-to-Code Comparison of Inter Lab Test Problem 1 for Asteroid Impact Hazard Mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Robert P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Howley, Kirsten [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferguson, Jim Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gisler, Galen Ross [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Plesko, Catherine Suzanne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Managan, Rob [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Owen, Mike [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wasem, Joseph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bruck-Syal, Megan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-15

    The NNSA Laboratories have entered into an interagency collaboration with the National Aeronautics and Space Administration (NASA) to explore strategies for prevention of Earth impacts by asteroids. Assessment of such strategies relies upon use of sophisticated multi-physics simulation codes. This document describes the task of verifying and cross-validating, between Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL), modeling capabilities and methods to be employed as part of the NNSA-NASA collaboration. The approach has been to develop a set of test problems and then to compare and contrast results obtained by use of a suite of codes, including MCNP, RAGE, Mercury, Ares, and Spheral. This document provides a short description of the codes, an overview of the idealized test problems, and discussion of the results for deflection by kinetic impactors and stand-off nuclear explosions.

  8. The data requirements for the verification and validation of a fuel performance code - the transuranus perspective

    International Nuclear Information System (INIS)

    Schubert, A.; Di Marcello, V.; Rondinella, V.; Van De Laar, J.; Van Uffelen, P.

    2013-01-01

    In general, the verification and validation (V and V) of a fuel performance code like TRANSURANUS consists of three basic steps: a) verifying the correctness and numerical stability of the sub-models; b) comparing the sub-models with experimental data; c) comparing the results of the integral fuel performance code with experimental data Only the second and third steps of the V and V rely on experimental information. This scheme can be further detailed according to the physical origin of the data: on one hand, in-reactor ('in-pile') experimental data are generated in the course of the irradiation; on the other hand ex-reactor ('out-of-pile') experimental data are obtained for instance from various postirradiation examinations (PIE) or dedicated experiments with fresh samples. For both categories, we will first discuss the V and V of sub-models of TRANSURANUS related to separate aspects of the fuel behaviour: this includes the radial variation of the composition and fissile isotopes, the thermal properties of the fuel (e.g. thermal conductivity, melting temperature, etc.), the mechanical properties of fuel and cladding (e.g. elastic constants, creep properties), as well as the models for the fission product behaviour. Secondly, the integral code verification will be addressed as it treats various aspects of the fuel behaviour, including the geometrical changes in the fuel and the gas pressure and composition of the free volume in the rod. (authors)

  9. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  10. Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

    International Nuclear Information System (INIS)

    Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

    2004-01-01

    Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

  11. Threats to Validity When Using Open-Ended Items in International Achievement Studies: Coding Responses to the PISA 2012 Problem-Solving Test in Finland

    Science.gov (United States)

    Arffman, Inga

    2016-01-01

    Open-ended (OE) items are widely used to gather data on student performance in international achievement studies. However, several factors may threaten validity when using such items. This study examined Finnish coders' opinions about threats to validity when coding responses to OE items in the PISA 2012 problem-solving test. A total of 6…

  12. The Analysis and the Performance Simulation of the Capacity of Bit-interleaved Coded Modulation System

    Directory of Open Access Journals (Sweden)

    Hongwei ZHAO

    2014-09-01

    Full Text Available In this paper, the capacity of the BICM system over AWGN channels is first analyzed; the curves of BICM capacity versus SNR are also got by the Monte-Carlo simulations===?=== and compared with the curves of the CM capacity. Based on the analysis results, we simulate the error performances of BICM system with LDPC codes. Simulation results show that the capacity of BICM system with LDPC codes is enormously influenced by the mapping methods. Given a certain modulation method, the BICM system can obtain about 2-3 dB gain with Gray mapping compared with Non-Gray mapping. Meanwhile, the simulation results also demonstrate the correctness of the theory analysis.

  13. Analyses with the FSTATE code: fuel performance in destructive in-pile experiments

    International Nuclear Information System (INIS)

    Bauer, T.H.; Meek, C.C.

    1982-01-01

    Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and application is made to in-pile experiments undertaken to study fast-reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss-of-cooling accident sequence, are used as representative examples, and the interpretation of STATE computations relative to experimental observations is made

  14. A Linear Algebra Framework for Static High Performance Fortran Code Distribution

    Directory of Open Access Journals (Sweden)

    Corinne Ancourt

    1997-01-01

    Full Text Available High Performance Fortran (HPF was developed to support data parallel programming for single-instruction multiple-data (SIMD and multiple-instruction multiple-data (MIMD machines with distributed memory. The programmer is provided a familiar uniform logical address space and specifies the data distribution by directives. The compiler then exploits these directives to allocate arrays in the local memories, to assign computations to elementary processors, and to migrate data between processors when required. We show here that linear algebra is a powerful framework to encode HPF directives and to synthesize distributed code with space-efficient array allocation, tight loop bounds, and vectorized communications for INDEPENDENT loops. The generated code includes traditional optimizations such as guard elimination, message vectorization and aggregation, and overlap analysis. The systematic use of an affine framework makes it possible to prove the compilation scheme correct.

  15. Achievements in testing of the MGA and FRAM isotopic software codes under the DOE/NNSA-IRSN cooperation of gamma-ray isotopic measurement systems

    International Nuclear Information System (INIS)

    Vo, Duc; Wang, Tzu-Fang; Funk, Pierre; Weber, Anne-Laure; Pepin, Nicolas; Karcher, Anna

    2009-01-01

    DOE/NNSA and IRSN collaborated on a study of gamma-ray instruments and analysis methods used to perform isotopic measurements of special nuclear materials. The two agencies agreed to collaborate on the project in response to inconsistencies that were found in the various versions of software and hardware used to determine the isotopic abundances of uranium and plutonium. IRSN used software developed internally to test the MGA and FRAM isotopic analysis codes for criteria used to stop data acquisition. The stop-criterion test revealed several unusual behaviors in both the MGA and FRAM software codes.

  16. COMPUTATION FORMAT computer codes X4TOC4 and PLOTC4. Implementing and Testing on a Personal Computer

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1987-05-01

    This document describes the contents of the diskette containing the COMPUTATION FORMAT codes X4TOC4 and PLOTC4 by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a single diskette. (author)

  17. Stand-alone containment analysis of Phébus FPT tests with ASTEC and MELCOR codes: the FPT-2 test.

    Science.gov (United States)

    Gonfiotti, Bruno; Paci, Sandro

    2018-03-01

    During the last 40 years, many studies have been carried out to investigate the different phenomena occurring during a Severe Accident (SA) in a Nuclear Power Plant (NPP). Such efforts have been supported by the execution of different experimental campaigns, and the integral Phébus FP tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR) fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus FP campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel (CV) have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FP) behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FP and transuranic elements from a debris bed and a pool of melted fuel.

  18. Stand-alone containment analysis of Phébus FPT tests with ASTEC and MELCOR codes: the FPT-2 test

    Directory of Open Access Journals (Sweden)

    Bruno Gonfiotti

    2018-03-01

    Full Text Available During the last 40 years, many studies have been carried out to investigate the different phenomena occurring during a Severe Accident (SA in a Nuclear Power Plant (NPP. Such efforts have been supported by the execution of different experimental campaigns, and the integral Phébus FP tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus FP campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel (CV have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FP behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FP and transuranic elements from a debris bed and a pool of melted fuel. Keywords: Safety

  19. An overview of the geochemical code MINTEQ: Applications to performance assessment for low-level wastes

    International Nuclear Information System (INIS)

    Peterson, S.R.; Opitz, B.E.; Graham, M.J.; Eary, L.E.

    1987-03-01

    The MINTEQ geochemical computer code, developed at the Pacific Northwest Laboratory (PNL), integrates many of the capabilities of its two immediate predecessors, MINEQL and WATEQ3. The MINTEQ code will be used in the Special Waste Form Lysimeters-Arid program to perform the calculations necessary to simulate (model) the contact of low-level waste solutions with heterogeneous sediments of the interaction of ground water with solidified low-level wastes. The code can calculate ion speciation/solubilitya, adsorption, oxidation-reduction, gas phase equilibria, and precipitation/dissolution of solid phases. Under the Special Waste Form Lysimeters-Arid program, the composition of effluents (leachates) from column and batch experiments, using laboratory-scale waste forms, will be used to develop a geochemical model of the interaction of ground water with commercial, solidified low-level wastes. The wastes being evaluated include power-reactor waste streams that have been solidified in cement, vinyl ester-styrene, and bitumen. The thermodynamic database for the code was upgraded preparatory to performing the geochemical modeling. Thermodynamic data for solid phases and aqueous species containing Sb, Ce, Cs, or Co were added to the MINTEQ database. The need to add these data was identified from the characterization of the waste streams. The geochemical model developed from the laboratory data will then be applied to predict the release from a field-lysimeter facility that contains full-scale waste samples. The contaminant concentrations migrating from the waste forms predicted using MINTEQ will be compared to the long-term lysimeter data. This comparison will constitute a partial field validation of the geochemical model

  20. Performance awareness execution performance of HEP codes on RISC platforms,issues and solutions

    CERN Document Server

    Yaari, R; Yaari, Refael; Jarp, Sverre

    1995-01-01

    The work described in this paper was started during the migration of Aleph's production jobs from the IBM mainframe/CRAY supercomputer to several RISC/Unix workstation platforms. The aim was to understand why Aleph did not obtain the performance on the RISC platforms that was "promised" after a CERN Unit comparison between these RISC platforms and the IBM mainframe. Remedies were also sought. Since the work with the Aleph jobs in turn led to the related task of understanding compilers and their options, the conditions under which the CERN benchmarks (and other benchmarks) were run, kernel routines and frequently used CERNLIB routines, the whole undertaking expanded to try to look at all the factors that influence the performance of High Energy Physics (HEP) jobs in general. Finally, key performance issues were reviewed against the programs of one of the LHC collaborations (Atlas) with the hope that the conclusions would be of long- term interest during the establishment of their simulation, reconstruction and...