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Sample records for specification code test

  1. Code, standard and specifications

    International Nuclear Information System (INIS)

    Abdul Nassir Ibrahim; Azali Muhammad; Ab. Razak Hamzah; Abd. Aziz Mohamed; Mohamad Pauzi Ismail

    2008-01-01

    Radiography also same as the other technique, it need standard. This standard was used widely and method of used it also regular. With that, radiography testing only practical based on regulations as mentioned and documented. These regulation or guideline documented in code, standard and specifications. In Malaysia, level one and basic radiographer can do radiography work based on instruction give by level two or three radiographer. This instruction was produced based on guideline that mention in document. Level two must follow the specifications mentioned in standard when write the instruction. From this scenario, it makes clearly that this radiography work is a type of work that everything must follow the rule. For the code, the radiography follow the code of American Society for Mechanical Engineer (ASME) and the only code that have in Malaysia for this time is rule that published by Atomic Energy Licensing Board (AELB) known as Practical code for radiation Protection in Industrial radiography. With the existence of this code, all the radiography must follow the rule or standard regulated automatically.

  2. 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

  3. Atlas C++ Coding Standard Specification

    CERN Document Server

    Albrand, S; Barberis, D; Bosman, M; Jones, B; Stavrianakou, M; Arnault, C; Candlin, D; Candlin, R; Franck, E; Hansl-Kozanecka, Traudl; Malon, D; Qian, S; Quarrie, D; Schaffer, R D

    2001-01-01

    This document defines the ATLAS C++ coding standard, that should be adhered to when writing C++ code. It has been adapted from the original "PST Coding Standard" document (http://pst.cern.ch/HandBookWorkBook/Handbook/Programming/programming.html) CERN-UCO/1999/207. The "ATLAS standard" comprises modifications, further justification and examples for some of the rules in the original PST document. All changes were discussed in the ATLAS Offline Software Quality Control Group and feedback from the collaboration was taken into account in the "current" version.

  4. 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

  5. Urine specific gravity test

    Science.gov (United States)

    ... medlineplus.gov/ency/article/003587.htm Urine specific gravity test To use the sharing features on this page, please enable JavaScript. Urine specific gravity is a laboratory test that shows the concentration ...

  6. Comparative Test Case Specification

    DEFF Research Database (Denmark)

    Kalyanova, Olena; Heiselberg, Per

     This document includes a definition of the comparative test cases DSF200_3 and DSF200_4, which previously described in the comparative test case specification for the test cases DSF100_3 and DSF200_3 [Ref.1]....... This document includes a definition of the comparative test cases DSF200_3 and DSF200_4, which previously described in the comparative test case specification for the test cases DSF100_3 and DSF200_3 [Ref.1]....

  7. Specification of a test problem for HYDROCOIN [Hydrologic Code Intercomparison] Level 3 Case 2: Sensitivity analysis for deep disposal in partially saturated, fractured tuff

    International Nuclear Information System (INIS)

    Prindle, R.W.

    1987-08-01

    The international Hydrologic Code Intercomparison Project (HYDROCOIN) was formed to evaluate hydrogeologic models and computer codes and their use in performance assessment for high-level radioactive waste repositories. Three principal activities in the HYDROCOIN Project are Level 1, verification and benchmarking of hydrologic codes; Level 2, validation of hydrologic models; and Level 3, sensitivity and uncertainty analyses of the models and codes. This report presents a test case defined for the HYDROCOIN Level 3 activity to explore the feasibility of applying various sensitivity-analysis methodologies to a highly nonlinear model of isothermal, partially saturated flow through fractured tuff, and to develop modeling approaches to implement the methodologies for sensitivity analysis. These analyses involve an idealized representation of a repository sited above the water table in a layered sequence of welded and nonwelded, fractured, volcanic tuffs. The analyses suggested here include one-dimensional, steady flow; one-dimensional, nonsteady flow; and two-dimensional, steady flow. Performance measures to be used to evaluate model sensitivities are also defined; the measures are related to regulatory criteria for containment of high-level radioactive waste. 14 refs., 5 figs., 4 tabs

  8. Comparative Test Case Specification

    DEFF Research Database (Denmark)

    Kalyanova, Olena; Heiselberg, Per

    This document includes the specification on the IEA task of evaluation building energy simulation computer programs for the Double Skin Facades (DSF) constructions. There are two approaches involved into this procedure, one is the comparative approach and another is the empirical one. In the comp....... In the comparative approach the outcomes of different software tools are compared, while in the empirical approach the modelling results are compared with the results of experimental test cases. The comparative test cases include: ventilation, shading and geometry....

  9. High efficiency video coding coding tools and specification

    CERN Document Server

    Wien, Mathias

    2015-01-01

    The video coding standard High Efficiency Video Coding (HEVC) targets at improved compression performance for video resolutions of HD and beyond, providing Ultra HD video at similar compressed bit rates as for HD video encoded with the well-established video coding standard H.264 | AVC. Based on known concepts, new coding structures and improved coding tools have been developed and specified in HEVC. The standard is expected to be taken up easily by established industry as well as new endeavors, answering the needs of todays connected and ever-evolving online world. This book presents the High Efficiency Video Coding standard and explains it in a clear and coherent language. It provides a comprehensive and consistently written description, all of a piece. The book targets at both, newbies to video coding as well as experts in the field. While providing sections with introductory text for the beginner, it suits as a well-arranged reference book for the expert. The book provides a comprehensive reference for th...

  10. Empirical Test Case Specification

    DEFF Research Database (Denmark)

    Kalyanova, Olena; Heiselberg, Per

    This document includes the empirical specification on the IEA task of evaluation building energy simulation computer programs for the Double Skin Facades (DSF) constructions. There are two approaches involved into this procedure, one is the comparative approach and another is the empirical one. I....... In the comparative approach the outcomes of different software tools are compared, while in the empirical approach the modelling results are compared with the results of experimental test cases....

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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

  16. Optimization Specifications for CUDA Code Restructuring Tool

    KAUST Repository

    Khan, Ayaz

    2017-03-13

    In this work we have developed a restructuring software tool (RT-CUDA) following the proposed optimization specifications to bridge the gap between high-level languages and the machine dependent CUDA environment. RT-CUDA takes a C program and convert it into an optimized CUDA kernel with user directives in a configuration file for guiding the compiler. RTCUDA also allows transparent invocation of the most optimized external math libraries like cuSparse and cuBLAS enabling efficient design of linear algebra solvers. We expect RT-CUDA to be needed by many KSA industries dealing with science and engineering simulation on massively parallel computers like NVIDIA GPUs.

  17. 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

  18. Hominoid-specific de novo protein-coding genes originating from long non-coding RNAs.

    Directory of Open Access Journals (Sweden)

    Chen Xie

    2012-09-01

    Full Text Available Tinkering with pre-existing genes has long been known as a major way to create new genes. Recently, however, motherless protein-coding genes have been found to have emerged de novo from ancestral non-coding DNAs. How these genes originated is not well addressed to date. Here we identified 24 hominoid-specific de novo protein-coding genes with precise origination timing in vertebrate phylogeny. Strand-specific RNA-Seq analyses were performed in five rhesus macaque tissues (liver, prefrontal cortex, skeletal muscle, adipose, and testis, which were then integrated with public transcriptome data from human, chimpanzee, and rhesus macaque. On the basis of comparing the RNA expression profiles in the three species, we found that most of the hominoid-specific de novo protein-coding genes encoded polyadenylated non-coding RNAs in rhesus macaque or chimpanzee with a similar transcript structure and correlated tissue expression profile. According to the rule of parsimony, the majority of these hominoid-specific de novo protein-coding genes appear to have acquired a regulated transcript structure and expression profile before acquiring coding potential. Interestingly, although the expression profile was largely correlated, the coding genes in human often showed higher transcriptional abundance than their non-coding counterparts in rhesus macaque. The major findings we report in this manuscript are robust and insensitive to the parameters used in the identification and analysis of de novo genes. Our results suggest that at least a portion of long non-coding RNAs, especially those with active and regulated transcription, may serve as a birth pool for protein-coding genes, which are then further optimized at the transcriptional level.

  19. Mother code specifications (Appendix to CEA report 2472)

    International Nuclear Information System (INIS)

    Pillard, Denise; Soule, Jean-Louis

    1964-12-01

    The Mother code (written in Fortran for IBM 7094) computes the integral cross section and the first two moments of energy transfer of a thermalizer. Computation organisation and methods are presented in an other document. This document presents code specifications, i.e. input data (for spectrum description, printing options, input record formats, conditions to be met by values), and results (printing formats and options, writing and punching options and formats)

  20. 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.

  1. 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

  2. Testing of badminton specific endurance

    DEFF Research Database (Denmark)

    Madsen, Christian Møller; Højlyng, Mads; Nybo, Lars

    2016-01-01

    In the present study, a novel intermittent badminton endurance test (B-ENDURANCE) was developed and tested in elite (n=17) and skilled (n=9) badminton players as well as in age-matched physically active men (non-badminton players; n=8). In addition, B-ENDURANCE test-retest reproducibility...... was evaluated in nine badminton players.B-ENDURANCE is an incremental test where each level consists of repeated sequences of badminton specific actions towards the four corners on the court. The subject starts in the center of the court in front of a computer screen and within each sequence he must...... decreases until the subjects cannot follow the dictated tempo.B-ENDURANCE performance for elite players was better (Pbadminton players. In addition, B-ENDURANCE performance correlated (r=0.8; P

  3. Interoperable domain-specific languages families for code generation

    Czech Academy of Sciences Publication Activity Database

    Malohlava, M.; Plášil, F.; Bureš, Tomáš; Hnětynka, P.

    2013-01-01

    Roč. 43, č. 5 (2013), s. 479-499 ISSN 0038-0644 R&D Projects: GA ČR GD201/09/H057 EU Projects: European Commission(XE) ASCENS 257414 Grant - others:GA AV ČR(CZ) GAP103/11/1489 Program:FP7 Institutional research plan: CEZ:AV0Z10300504 Keywords : code generation * domain specific languages * models reuse * extensible languages * specification * program synthesis Subject RIV: JC - Computer Hardware ; Software Impact factor: 1.148, year: 2013

  4. Site-specific Probabilistic Analysis of DCGLs Using RESRAD Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeongju; Yoon, Suk Bon; Sohn, Wook [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    In general, DCGLs can be conservative (screening DCGL) if they do not take into account site specific factors. Use of such conservative DCGLs can lead to additional remediation that would not be required if the effort was made to develop site-specific DCGLs. Therefore, the objective of this work is to provide an example on the use of the RESRAD 6.0 probabilistic (site-specific) dose analysis to compare with the screening DCGL. Site release regulations state that a site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group of less than the site release criteria, for example 0.25 mSv per year in U.S. Utilities use computer dose modeling codes to establish an acceptable level of contamination, the derived concentration guideline level (DCGL) that will meet this regulatory limit. Since the DCGL value is the principal measure of residual radioactivity, it is critical to understand the technical basis of these dose modeling codes. The objective this work was to provide example on nuclear power plant decommissioning dose analysis in a probabilistic analysis framework. The focus was on the demonstration of regulatory compliance for surface soil contamination using the RESRAD 6.0 code. Both the screening and site-specific probabilistic dose analysis methodologies were examined. Example analyses performed with the screening probabilistic dose analysis confirmed the conservatism of the NRC screening values and indicated the effectiveness of probabilistic dose analysis in reducing the conservatism in DCGL derivation.

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. Experiment-specific analyses in support of code development

    International Nuclear Information System (INIS)

    Ott, L.J.

    1990-01-01

    Experiment-specific models have been developed since 1986 by Oak Ridge National Laboratory Boiling Water Reactor (BWR) severe accident analysis programs for the purpose of BWR experimental planning and optimum interpretation of experimental results. These experiment-specific models have been applied to large integral tests (ergo, experiments) which start from an initial undamaged core state. The tests performed to date in BWR geometry have had significantly different-from-prototypic boundary and experimental conditions because of either normal facility limitations or specific experimental constraints. These experiments (ACRR: DF-4, NRU: FLHT-6, and CORA) were designed to obtain specific phenomenological information such as the degradation and interaction of prototypic components and the effects on melt progression of control-blade materials and channel boxes. Applications of ORNL models specific to the ACRR DF-4 and KfK CORA-16 experiments are discussed and significant findings from the experimental analyses are presented. 32 refs., 16 figs

  10. Remote Testing of Timed Specifications

    DEFF Research Database (Denmark)

    David, Alexandre; Larsen, Kim Guldstrand; Mikucionis, Marius

    2013-01-01

    We present a study and a testing framework on black box remote testing of real-time systems using UPPAAL TIGA. One of the essential challenges of remote testing is the communication latency between the Tester and the System Under Test (IUT) that may lead to interleaving of inputs and outputs. Thi...

  11. 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)

  12. 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.

  13. YOUTH BASKETBALL SPECIFIC EFFORT TEST

    Directory of Open Access Journals (Sweden)

    Philippe Campillo

    2004-12-01

    Full Text Available The test, as it is presented, must be modified to produce more pertinent results. The measure before the maximum jumps in RJ can also show at what intensity the athletes produce the jump repetitions. The knowledge of the maximal performance during a jump constitutes a reference with which one can evaluate a subject's commitment and efficiency during the test. The results of the study show that this type of test protocol can be a good method to evaluate the physical condition of an athlete during training.

  14. 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

  15. 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.

  16. 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

  17. Final Empirical Test Case Specification

    DEFF Research Database (Denmark)

    Kalyanova, Olena; Heiselberg, Per

    This document includes the empirical specification on the IEA task of evaluation building energy simulation computer programs for the Double Skin Facades (DSF) constructions. There are two approaches involved into this procedure, one is the comparative approach and another is the empirical one....

  18. 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

  19. 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

  20. Correlated sampling added to the specific purpose Monte Carlo code McPNL for neutron lifetime log responses

    International Nuclear Information System (INIS)

    Mickael, M.; Verghese, K.; Gardner, R.P.

    1989-01-01

    The specific purpose neutron lifetime oil well logging simulation code, McPNL, has been rewritten for greater user-friendliness and faster execution. Correlated sampling has been added to the code to enable studies of relative changes in the tool response caused by environmental changes. The absolute responses calculated by the code have been benchmarked against laboratory test pit data. The relative responses from correlated sampling are not directly benchmarked, but they are validated using experimental and theoretical results

  1. 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.)

  2. Technical specification of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y

    1998-03-01

    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 operation limit, safety limit, operation condition and checking points of HANARO fuel test loop. This results will become guidances for the planning of irradiation testing and operation of HANARO fuel test loop. (author). 13 refs., 13 tabs., 8 figs.

  3. A regulatory code for neuron-specific odor receptor expression.

    Directory of Open Access Journals (Sweden)

    Anandasankar Ray

    2008-05-01

    Full Text Available Olfactory receptor neurons (ORNs must select-from a large repertoire-which odor receptors to express. In Drosophila, most ORNs express one of 60 Or genes, and most Or genes are expressed in a single ORN class in a process that produces a stereotyped receptor-to-neuron map. The construction of this map poses a problem of receptor gene regulation that is remarkable in its dimension and about which little is known. By using a phylogenetic approach and the genome sequences of 12 Drosophila species, we systematically identified regulatory elements that are evolutionarily conserved and specific for individual Or genes of the maxillary palp. Genetic analysis of these elements supports a model in which each receptor gene contains a zip code, consisting of elements that act positively to promote expression in a subset of ORN classes, and elements that restrict expression to a single ORN class. We identified a transcription factor, Scalloped, that mediates repression. Some elements are used in other chemosensory organs, and some are conserved upstream of axon-guidance genes. Surprisingly, the odor response spectra and organization of maxillary palp ORNs have been extremely well-conserved for tens of millions of years, even though the amino acid sequences of the receptors are not highly conserved. These results, taken together, define the logic by which individual ORNs in the maxillary palp select which odor receptors to express.

  4. 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.

  5. Software requirements specification document for the AREST code development

    International Nuclear Information System (INIS)

    Engel, D.W.; McGrail, B.P.; Whitney, P.D.; Gray, W.J.; Williford, R.E.; White, M.D.; Eslinger, P.W.; Altenhofen, M.K.

    1993-11-01

    The Analysis of the Repository Source Term (AREST) computer code was selected in 1992 by the U.S. Department of Energy. The AREST code will be used to analyze the performance of an underground high level nuclear waste repository. The AREST code is being modified by the Pacific Northwest Laboratory (PNL) in order to evaluate the engineered barrier and waste package designs, model regulatory compliance, analyze sensitivities, and support total systems performance assessment modeling. The current version of the AREST code was developed to be a very useful tool for analyzing model uncertainties and sensitivities to input parameters. The code has also been used successfully in supplying source-terms that were used in a total systems performance assessment. The current version, however, has been found to be inadequate for the comparison and selection of a design for the waste package. This is due to the assumptions and simplifications made in the selection of the process and system models. Thus, the new version of the AREST code will be designed to focus on the details of the individual processes and implementation of more realistic models. This document describes the requirements of the new models that will be implemented. Included in this document is a section describing the near-field environmental conditions for this waste package modeling, description of the new process models that will be implemented, and a description of the computer requirements for the new version of the AREST code

  6. 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.

  7. 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

  8. Design specifications for ASME B and PV Code Section III nuclear class 1 piping

    International Nuclear Information System (INIS)

    Richardson, J.A.

    1978-01-01

    ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

  9. Prostate-Specific Antigen (PSA) Test

    Science.gov (United States)

    ... Cancer Prostate Cancer Screening Research Prostate-Specific Antigen (PSA) Test On This Page What is the PSA ... parts of the body before being detected. The PSA test may give false-positive or false-negative ...

  10. Code Shift: Grid Specifications and Dynamic Wind Turbine Models

    DEFF Research Database (Denmark)

    Ackermann, Thomas; Ellis, Abraham; Fortmann, Jens

    2013-01-01

    Grid codes (GCs) and dynamic wind turbine (WT) models are key tools to allow increasing renewable energy penetration without challenging security of supply. In this article, the state of the art and the further development of both tools are discussed, focusing on the European and North American e...

  11. 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.

  12. 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

  13. 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)

  14. Project NEO Specific Impulse Testing Solutions

    Science.gov (United States)

    Baffa, Bill

    2018-01-01

    The Neo test stand is currently configured to fire a horizontally mounted rocket motor with up to 6500 lbf thrust. Currently, the Neo test stand can measure flow of liquid propellant and oxidizer, pressures residing in the closed system up to the combustion chamber. The current configuration does not have the ability to provide all data needed to compute specific impulse. This presents three methods to outfit the NEO test fixture with instrumentation allowing for calculation of specific impulse.

  15. 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

  16. Generating Safety-Critical PLC Code From a High-Level Application Software Specification

    Science.gov (United States)

    2008-01-01

    The benefits of automatic-application code generation are widely accepted within the software engineering community. These benefits include raised abstraction level of application programming, shorter product development time, lower maintenance costs, and increased code quality and consistency. Surprisingly, code generation concepts have not yet found wide acceptance and use in the field of programmable logic controller (PLC) software development. Software engineers at Kennedy Space Center recognized the need for PLC code generation while developing the new ground checkout and launch processing system, called the Launch Control System (LCS). Engineers developed a process and a prototype software tool that automatically translates a high-level representation or specification of application software into ladder logic that executes on a PLC. All the computer hardware in the LCS is planned to be commercial off the shelf (COTS), including industrial controllers or PLCs that are connected to the sensors and end items out in the field. Most of the software in LCS is also planned to be COTS, with only small adapter software modules that must be developed in order to interface between the various COTS software products. A domain-specific language (DSL) is a programming language designed to perform tasks and to solve problems in a particular domain, such as ground processing of launch vehicles. The LCS engineers created a DSL for developing test sequences of ground checkout and launch operations of future launch vehicle and spacecraft elements, and they are developing a tabular specification format that uses the DSL keywords and functions familiar to the ground and flight system users. The tabular specification format, or tabular spec, allows most ground and flight system users to document how the application software is intended to function and requires little or no software programming knowledge or experience. A small sample from a prototype tabular spec application is

  17. 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

  18. 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.

  19. 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.)

  20. 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.

  1. 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

  2. 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

  3. Testing of Badminton-Specific Endurance.

    Science.gov (United States)

    Madsen, Christian M; Højlyng, Mads; Nybo, Lars

    2016-09-01

    Madsen, CM, Højlyng, M, and Nybo, L. Testing of badminton-specific endurance. J Strength Cond Res 30(9): 2582-2590, 2016-In the present study, a novel intermittent badminton endurance (B-ENDURANCE) test was developed and tested in elite (n = 17) and skilled (n = 9) badminton players and in age-matched physically active men (nonbadminton players; n = 8). In addition, B-ENDURANCE test-retest reproducibility was evaluated in 9 badminton players. The B-ENDURANCE test is an incremental test where each level consists of repeated sequences of badminton-specific actions toward the 4 corners of the court. The subject starts in the center of the court in front of a computer screen and within each sequence, he must, in a randomized order, complete 8 actions as dictated by the computer, providing the audiovisual input and verifying that the appropriate sensor is activated within the allocated time. Recovery time between each sequence is 10 seconds throughout the test, but the time to complete each sequence is gradually decreased until the subjects cannot follow the dictated tempo. The B-ENDURANCE test performance for elite players was better (p ≤ 0.05) compared with the skilled players and nonbadminton players. In addition, the B-ENDURANCE test performance correlated (r = 0.8 and p badminton-specific endurance but at least 1 familiarization trial is recommended if the test is used for evaluation of longitudinal changes, e.g., tracking training effects.

  4. NIF small optics laser damage test specifications

    International Nuclear Information System (INIS)

    Sheehan, L

    1999-01-01

    The Laser Damage Group is currently conducting tests on small optics samples supplied for initial evaluation of potential NIF suppliers. This document is meant to define the specification of laser-induced damage for small optics and the test methods used to collect the data. A rating system which will be applied for vendor selection is presented

  5. 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

  6. 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)

  7. 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

  8. Surface-specific additive manufacturing test artefacts

    Science.gov (United States)

    Townsend, Andrew; Racasan, Radu; Blunt, Liam

    2018-06-01

    Many test artefact designs have been proposed for use with additive manufacturing (AM) systems. These test artefacts have primarily been designed for the evaluation of AM form and dimensional performance. A series of surface-specific measurement test artefacts designed for use in the verification of AM manufacturing processes are proposed here. Surface-specific test artefacts can be made more compact because they do not require the large dimensions needed for accurate dimensional and form measurements. The series of three test artefacts are designed to provide comprehensive information pertaining to the manufactured surface. Measurement possibilities include deviation analysis, surface texture parameter data generation, sub-surface analysis, layer step analysis and build resolution comparison. The test artefacts are designed to provide easy access for measurement using conventional surface measurement techniques, for example, focus variation microscopy, stylus profilometry, confocal microscopy and scanning electron microscopy. Additionally, the test artefacts may be simply visually inspected as a comparative tool, giving a fast indication of process variation between builds. The three test artefacts are small enough to be included in every build and include built-in manufacturing traceability information, making them a convenient physical record of the build.

  9. Domain-specific modeling enabling full code generation

    CERN Document Server

    Kelly, Steven

    2007-01-01

    Domain-Specific Modeling (DSM) is the latest approach tosoftware development, promising to greatly increase the speed andease of software creation. Early adopters of DSM have been enjoyingproductivity increases of 500–1000% in production for over adecade. This book introduces DSM and offers examples from variousfields to illustrate to experienced developers how DSM can improvesoftware development in their teams. Two authorities in the field explain what DSM is, why it works,and how to successfully create and use a DSM solution to improveproductivity and quality. Divided into four parts, the book covers:background and motivation; fundamentals; in-depth examples; andcreating DSM solutions. There is an emphasis throughout the book onpractical guidelines for implementing DSM, including how toidentify the nece sary language constructs, how to generate fullcode from models, and how to provide tool support for a new DSMlanguage. The example cases described in the book are available thebook's Website, www.dsmbook....

  10. Analysis of genetic code ambiguity arising from nematode-specific misacylated tRNAs.

    Directory of Open Access Journals (Sweden)

    Kiyofumi Hamashima

    Full Text Available The faithful translation of the genetic code requires the highly accurate aminoacylation of transfer RNAs (tRNAs. However, it has been shown that nematode-specific V-arm-containing tRNAs (nev-tRNAs are misacylated with leucine in vitro in a manner that transgresses the genetic code. nev-tRNA(Gly (CCC and nev-tRNA(Ile (UAU, which are the major nev-tRNA isotypes, could theoretically decode the glycine (GGG codon and isoleucine (AUA codon as leucine, causing GGG and AUA codon ambiguity in nematode cells. To test this hypothesis, we investigated the functionality of nev-tRNAs and their impact on the proteome of Caenorhabditis elegans. Analysis of the nucleotide sequences in the 3' end regions of the nev-tRNAs showed that they had matured correctly, with the addition of CCA, which is a crucial posttranscriptional modification required for tRNA aminoacylation. The nuclear export of nev-tRNAs was confirmed with an analysis of their subcellular localization. These results show that nev-tRNAs are processed to their mature forms like common tRNAs and are available for translation. However, a whole-cell proteome analysis found no detectable level of nev-tRNA-induced mistranslation in C. elegans cells, suggesting that the genetic code is not ambiguous, at least under normal growth conditions. Our findings indicate that the translational fidelity of the nematode genetic code is strictly maintained, contrary to our expectations, although deviant tRNAs with misacylation properties are highly conserved in the nematode genome.

  11. Application of software quality assurance to a specific scientific code development task

    International Nuclear Information System (INIS)

    Dronkers, J.J.

    1986-03-01

    This paper describes an application of software quality assurance to a specific scientific code development program. The software quality assurance program consists of three major components: administrative control, configuration management, and user documentation. The program attempts to be consistent with existing local traditions of scientific code development while at the same time providing a controlled process of development

  12. 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.

  13. 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...

  14. 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.

  15. 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

  16. Tests in Print II: An Index to Tests, Test Reviews, and the Literature on Specific Tests.

    Science.gov (United States)

    Buros, Oscar K., Ed.

    Tests in Print II is a comprehensive, annotated bibliography of all in-print tests published as separates for use with English-speaking subjects. The 1,155 two-column pages list 2,467 tests in print as of early 1974; 16,574 references through 1971 on specific tests; a reprinting of the 1974 APA-AERA-NCME Standards for Educational andPsychological…

  17. Disease-Specific Trends of Comorbidity Coding and Implications for Risk Adjustment in Hospital Administrative Data.

    Science.gov (United States)

    Nimptsch, Ulrike

    2016-06-01

    To investigate changes in comorbidity coding after the introduction of diagnosis related groups (DRGs) based prospective payment and whether trends differ regarding specific comorbidities. Nationwide administrative data (DRG statistics) from German acute care hospitals from 2005 to 2012. Observational study to analyze trends in comorbidity coding in patients hospitalized for common primary diseases and the effects on comorbidity-related risk of in-hospital death. Comorbidity coding was operationalized by Elixhauser diagnosis groups. The analyses focused on adult patients hospitalized for the primary diseases of heart failure, stroke, and pneumonia, as well as hip fracture. When focusing the total frequency of diagnosis groups per record, an increase in depth of coding was observed. Between-hospital variations in depth of coding were present throughout the observation period. Specific comorbidity increases were observed in 15 of the 31 diagnosis groups, and decreases in comorbidity were observed for 11 groups. In patients hospitalized for heart failure, shifts of comorbidity-related risk of in-hospital death occurred in nine diagnosis groups, in which eight groups were directed toward the null. Comorbidity-adjusted outcomes in longitudinal administrative data analyses may be biased by nonconstant risk over time, changes in completeness of coding, and between-hospital variations in coding. Accounting for such issues is important when the respective observation period coincides with changes in the reimbursement system or other conditions that are likely to alter clinical coding practice. © Health Research and Educational Trust.

  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. 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.

  20. RETRANS - A tool to verify the functional equivalence of automatically generated source code with its specification

    International Nuclear Information System (INIS)

    Miedl, H.

    1998-01-01

    Following the competent technical standards (e.g. IEC 880) it is necessary to verify each step in the development process of safety critical software. This holds also for the verification of automatically generated source code. To avoid human errors during this verification step and to limit the cost effort a tool should be used which is developed independently from the development of the code generator. For this purpose ISTec has developed the tool RETRANS which demonstrates the functional equivalence of automatically generated source code with its underlying specification. (author)

  1. 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.

  2. Utilization of genetic tests: analysis of gene-specific billing in Medicare claims data.

    Science.gov (United States)

    Lynch, Julie A; Berse, Brygida; Dotson, W David; Khoury, Muin J; Coomer, Nicole; Kautter, John

    2017-08-01

    We examined the utilization of precision medicine tests among Medicare beneficiaries through analysis of gene-specific tier 1 and 2 billing codes developed by the American Medical Association in 2012. We conducted a retrospective cross-sectional study. The primary source of data was 2013 Medicare 100% fee-for-service claims. We identified claims billed for each laboratory test, the number of patients tested, expenditures, and the diagnostic codes indicated for testing. We analyzed variations in testing by patient demographics and region of the country. Pharmacogenetic tests were billed most frequently, accounting for 48% of the expenditures for new codes. The most common indications for testing were breast cancer, long-term use of medications, and disorders of lipid metabolism. There was underutilization of guideline-recommended tumor mutation tests (e.g., epidermal growth factor receptor) and substantial overutilization of a test discouraged by guidelines (methylenetetrahydrofolate reductase). Methodology-based tier 2 codes represented 15% of all claims billed with the new codes. The highest rate of testing per beneficiary was in Mississippi and the lowest rate was in Alaska. Gene-specific billing codes significantly improved our ability to conduct population-level research of precision medicine. Analysis of these data in conjunction with clinical records should be conducted to validate findings.Genet Med advance online publication 26 January 2017.

  3. Multiloop Integral System Test (MIST): MIST Facility Functional Specification

    International Nuclear Information System (INIS)

    Habib, T.F.; Koksal, C.G.; Moskal, T.E.; Rush, G.C.; Gloudemans, J.R.

    1991-04-01

    The Multiloop Integral System Test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock ampersand Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facilities to address the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility -- the Once Through Integral System (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST Functional Specification documents as-built design features, dimensions, instrumentation, and test approach. It also presents the scaling basis for the facility and serves to define the scope of work for the facility design and construction. 13 refs., 112 figs., 38 tabs

  4. 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.

  5. 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

  6. 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

  7. The computer code SEURBNUK/EURDYN (Release 1). Input and output specification

    International Nuclear Information System (INIS)

    Broadhouse, B.J.; Yerkess, A.

    1986-05-01

    SEURBNUK/EURODYN is an extension of SEURBNUK-2, a two dimensional, axisymmetric, Eulerian, finite element containment code in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin and thick structures. These codes are designed to model the hydrodynamic development in time of a hypothetical core disruptive accident (HCDA) in a fast breeder reactor. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations, with information on output facilities, and aid to users to avoid some common difficulties. (UK)

  8. 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

  9. Task specificity of finger dexterity tests

    NARCIS (Netherlands)

    Berger, M.A.M.; Krul, A.; Daanen, H.A.M.

    2009-01-01

    Finger dexterity tests are generally used to assess performance decrease due to gloves, cold and pathology. It is generally assumed that the O’Connor and Purdue Pegboard test yield similar results. In this experiment we compared these two tests for dry conditions without gloves, and for dry and wet

  10. Task specificity of finger dexterity tests

    NARCIS (Netherlands)

    Berger, M.A.M.; Krul, A.J.; Daanen, H.A.M.

    2009-01-01

    Finger dexterity tests are generally used to assess performance decrease due to gloves, cold and pathology. It is generally assumed that the O'Connor and Purdue Pegboard test yield similar results. In this experiment we compared these two tests for dry conditions without gloves, and for dry and wet

  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. Code-specific learning rules improve action selection by populations of spiking neurons.

    Science.gov (United States)

    Friedrich, Johannes; Urbanczik, Robert; Senn, Walter

    2014-08-01

    Population coding is widely regarded as a key mechanism for achieving reliable behavioral decisions. We previously introduced reinforcement learning for population-based decision making by spiking neurons. Here we generalize population reinforcement learning to spike-based plasticity rules that take account of the postsynaptic neural code. We consider spike/no-spike, spike count and spike latency codes. The multi-valued and continuous-valued features in the postsynaptic code allow for a generalization of binary decision making to multi-valued decision making and continuous-valued action selection. We show that code-specific learning rules speed up learning both for the discrete classification and the continuous regression tasks. The suggested learning rules also speed up with increasing population size as opposed to standard reinforcement learning rules. Continuous action selection is further shown to explain realistic learning speeds in the Morris water maze. Finally, we introduce the concept of action perturbation as opposed to the classical weight- or node-perturbation as an exploration mechanism underlying reinforcement learning. Exploration in the action space greatly increases the speed of learning as compared to exploration in the neuron or weight space.

  13. 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

  14. Chemical cleaning specification: few tube test model

    International Nuclear Information System (INIS)

    Hampton, L.V.; Simpson, J.L.

    1979-09-01

    The specification is for the waterside chemical cleaning of the 2 1/4 Cr - 1 Mo steel steam generator tubes. It describes the reagents and conditions for post-chemical cleaning passivation of the evaporator tubes

  15. Assessment of RELAP5/Mod3 system thermal hydraulic code using power test data of a BWR6 reactor

    International Nuclear Information System (INIS)

    Lee, M.; Chiang, C.S.

    1997-01-01

    The power test data of Kuosheng Nuclear Power Plant were used to assess RELAP5/Mod3 system thermal hydraulic analysis code. The plant employed a General Electric designed Boiling Water Reactor (BWR6) with rated power of 2894 MWth. The purpose of the assessment is to verify the validity of the plant specific RELAP5/Mod3 input deck for transient analysis. The power tests considered in the assessment were 100% power generator load rejection, the closure of main steam isolation valves (MSIVs) at 96% power, and the trip of recirculation pumps at 68% power. The major parameters compared in the assessment were steam dome pressure, steam flow rate, core flow rate, and downcomer water level. The comparisons of the system responses predicted by the code and the power test data were reasonable which demonstrated the capabilities of the code and the validity of the input deck. However, it was also identified that the separator model of the code may cause energy imbalance problem in the transient calculation. In the assessment, the steam separators were modeled using time-dependent junctions. In the approach, a complete separation of steam and water was predicted. The system responses predicted by RELAP5/Mod3 code were also compared with those from the calculations of RETRAN code. When these results were compared with the power test data, the predictions of the RETRAN code were better than those of RELAP5/Mod3. In the simulation of 100% power generator load rejection, it was believed that the difference in the steam separator model of these two codes was one of the reason of the difference in the prediction of power test data. The predictions of RELAP/Mod3 code can also be improved by the incorporation of one-dimensional kinetic model. There was also some margin for the improvement of the input related to the feedwater control system. (author)

  16. 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

  17. 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

  18. L1 and L2 Picture Naming in Mandarin-English Bilinguals: A Test of Bilingual Dual Coding Theory

    Science.gov (United States)

    Jared, Debra; Poh, Rebecca Pei Yun; Paivio, Allan

    2013-01-01

    This study examined the nature of bilinguals' conceptual representations and the links from these representations to words in L1 and L2. Specifically, we tested an assumption of the Bilingual Dual Coding Theory that conceptual representations include image representations, and that learning two languages in separate contexts can result in…

  19. 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)

  20. 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.

  1. Design implications for task-specific search utilities for retrieval and re-engineering of code

    Science.gov (United States)

    Iqbal, Rahat; Grzywaczewski, Adam; Halloran, John; Doctor, Faiyaz; Iqbal, Kashif

    2017-05-01

    The importance of information retrieval systems is unquestionable in the modern society and both individuals as well as enterprises recognise the benefits of being able to find information effectively. Current code-focused information retrieval systems such as Google Code Search, Codeplex or Koders produce results based on specific keywords. However, these systems do not take into account developers' context such as development language, technology framework, goal of the project, project complexity and developer's domain expertise. They also impose additional cognitive burden on users in switching between different interfaces and clicking through to find the relevant code. Hence, they are not used by software developers. In this paper, we discuss how software engineers interact with information and general-purpose information retrieval systems (e.g. Google, Yahoo!) and investigate to what extent domain-specific search and recommendation utilities can be developed in order to support their work-related activities. In order to investigate this, we conducted a user study and found that software engineers followed many identifiable and repeatable work tasks and behaviours. These behaviours can be used to develop implicit relevance feedback-based systems based on the observed retention actions. Moreover, we discuss the implications for the development of task-specific search and collaborative recommendation utilities embedded with the Google standard search engine and Microsoft IntelliSense for retrieval and re-engineering of code. Based on implicit relevance feedback, we have implemented a prototype of the proposed collaborative recommendation system, which was evaluated in a controlled environment simulating the real-world situation of professional software engineers. The evaluation has achieved promising initial results on the precision and recall performance of the system.

  2. A human-specific de novo protein-coding gene associated with human brain functions.

    Directory of Open Access Journals (Sweden)

    Chuan-Yun Li

    2010-03-01

    Full Text Available To understand whether any human-specific new genes may be associated with human brain functions, we computationally screened the genetic vulnerable factors identified through Genome-Wide Association Studies and linkage analyses of nicotine addiction and found one human-specific de novo protein-coding gene, FLJ33706 (alternative gene symbol C20orf203. Cross-species analysis revealed interesting evolutionary paths of how this gene had originated from noncoding DNA sequences: insertion of repeat elements especially Alu contributed to the formation of the first coding exon and six standard splice junctions on the branch leading to humans and chimpanzees, and two subsequent substitutions in the human lineage escaped two stop codons and created an open reading frame of 194 amino acids. We experimentally verified FLJ33706's mRNA and protein expression in the brain. Real-Time PCR in multiple tissues demonstrated that FLJ33706 was most abundantly expressed in brain. Human polymorphism data suggested that FLJ33706 encodes a protein under purifying selection. A specifically designed antibody detected its protein expression across human cortex, cerebellum and midbrain. Immunohistochemistry study in normal human brain cortex revealed the localization of FLJ33706 protein in neurons. Elevated expressions of FLJ33706 were detected in Alzheimer's brain samples, suggesting the role of this novel gene in human-specific pathogenesis of Alzheimer's disease. FLJ33706 provided the strongest evidence so far that human-specific de novo genes can have protein-coding potential and differential protein expression, and be involved in human brain functions.

  3. Improvement of the test quality for specific test problems. Proceedings

    International Nuclear Information System (INIS)

    2011-01-01

    This proceedings CD discusses the many factors that are relevant in nearly all tests, as well as their effects on the validity of the test result. Interfaces with technical rules, staff qualification, POD, and validation of test results by supplementary techniques are presented as well. Three of the 17 papers are available as separate records in the ENERGY database. [de

  4. Test Specification of A1-1 Test for OECD-ATLAS Project

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Moon, Sang-Ki; Lee, Seung-Wook; Choi, Ki-Yong; Song, Chul-Hwa

    2014-01-01

    In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to meet the international interests in the multiple high-risk DECs raised after the Fukushima accident. The proposed test matrix for the OECD-ATLAS project is summarized in Table 1.. In this study, detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG number 2 (SG-2). A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project. A detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG-2 in order to consider an accident mitigation measure. The pre-test analysis using MARS code was performed with an aim of setting up the detailed test procedures for A1-1 test and also gaining the physical insights for a prolonged SBO transient. In the A1-1 test, a prolonged SBO transient will be simulated with two temporal phases: Phase (I) for conservative SBO transient without supply of turbine-driven auxiliary feedwater and Phase (II) for asymmetric cooling via single trained supply of turbine-driven auxiliary feedwater

  5. STEEP4 code for computation of specific thermonuclear reaction rates from pointwise cross sections

    International Nuclear Information System (INIS)

    Harris, D.R.; Dei, D.E.; Husseiny, A.A.; Sabri, Z.A.; Hale, G.M.

    1976-05-01

    A code module, STEEP4, is developed to calculate the fusion reaction rates in terms of the specific reactivity [sigma v] which is the product of cross section and relative velocity averaged over the actual ion distributions of the interacting particles in the plasma. The module is structured in a way suitable for incorporation in thermonuclear burn codes to provide rapid and yet relatively accurate on-line computation of [sigma v] as a function of plasma parameters. Ion distributions are modified to include slowing-down contributions which are characterized in terms of plasma parameters. Rapid and accurate algorithms are used for integrating [sigma v] from cross sections and spectra. The main program solves for [sigma v] by the method of steepest descent. However, options are provided to use Gauss-Hermite and dense trapezoidal quadrature integration techniques. Options are also provided for rapid calculation of screening effects on specific reaction rates. Although such effects are not significant in cases of plasmas of laboratory interest, the options are included to increase the range of applicability of the code. Gamow penetration form, log-log interpolation, and cubic interpolation routines are included to provide the interpolated values of cross sections

  6. Prostate-Specific Antigen (PSA) Test: MedlinePlus Lab Test Information

    Science.gov (United States)

    ... medlineplus.gov/labtests/prostatespecificantigenpsatest.html Prostate-Specific Antigen (PSA) Test To use the sharing features on this ... enable JavaScript. What is a prostate-specific antigen (PSA) test? A prostate-specific antigen (PSA) test measures ...

  7. 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

  8. 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.

  9. 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.

  10. 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)

  11. The computer code SEURBNUK/EURDYN (release 1). Input and output specifications

    International Nuclear Information System (INIS)

    Smith, B.L.; Broadhouse, B.J.; Yerkess, A.

    1986-05-01

    SEURBNUK-2 is a two-dimensional, axisymmetric, Eulerian, finite difference containment code developed initially by AWRE Aldermaston, AEE Winfrith and JRC-Ispra, and more recently by AEEW, JRC and EIR Wuerenlingen. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method which itself is an extension of the MAC algorithm. SEURBNUK has a finite difference thin shell treatment for vessels and internal structures of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. SEURBNUK/EURDYN is an extension of SEURBNUK-2 in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin or thick structures. This has been achieved by coupling the finite element code EURDYN with SEURBNUK-2, allowing the use of conical shell elements and axisymmetric triangular elements. Within the code, the equations of motion for the structures are solved quite separately from those for the fluid, and the timestep for the fluid can be an integer multiple of that for the structures. The interaction of the structures with the fluid is then considered as a modification to the coefficients in the pressure equations, the modifications naturally depending on the behaviour of the structures within the fluid cell. The code is limited to dealing with a single fluid, the coolant, and the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations. After explaining the output facilities information is included to aid users to avoid some common pit-falls. (author)

  12. 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

  13. 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.

  14. 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

  15. 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…

  16. 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…

  17. The PP1 binding code: a molecular-lego strategy that governs specificity.

    Science.gov (United States)

    Heroes, Ewald; Lesage, Bart; Görnemann, Janina; Beullens, Monique; Van Meervelt, Luc; Bollen, Mathieu

    2013-01-01

    Ser/Thr protein phosphatase 1 (PP1) is a single-domain hub protein with nearly 200 validated interactors in vertebrates. PP1-interacting proteins (PIPs) are ubiquitously expressed but show an exceptional diversity in brain, testis and white blood cells. The binding of PIPs is mainly mediated by short motifs that dock to surface grooves of PP1. Although PIPs often contain variants of the same PP1 binding motifs, they differ in the number and combination of docking sites. This molecular-lego strategy for binding to PP1 creates holoenzymes with unique properties. The PP1 binding code can be described as specific, universal, degenerate, nonexclusive and dynamic. PIPs control associated PP1 by interference with substrate recruitment or access to the active site. In addition, some PIPs have a subcellular targeting domain that promotes dephosphorylation by increasing the local concentration of PP1. The diversity of the PP1 interactome and the properties of the PP1 binding code account for the exquisite specificity of PP1 in vivo. © 2012 The Authors Journal compilation © 2012 FEBS.

  18. 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

  19. 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

  20. 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.

  1. Specifications, tests, and installation of wires and cables for the Diablo Canyon Nuclear Power Project

    International Nuclear Information System (INIS)

    Dan, F.J.

    1977-01-01

    The process of selecting wires and cables for the Diablo Canyon Nuclear Power Project is described. The criteria for the fire and environmental tests, the basis for the specifications, and the reasons for the final choice and acceptance are outlined. A short section is dedicated to the installation of cables in raceways with reference to separation and color coding. Also covered are the selection and testing of fire stops and the selection of seismic supports

  2. A comprehensive test specification for pulse fission counters

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, D L [Control and Instrumentation Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1962-02-15

    The following test specification is based on the memorandum AERE - M 728 which it now replaces It contains a standard acceptance test procedure for the many U.K.A.E.A, designed pulse fission counters now commercially available. This test specification may be used for any pulse fission counter provided a specification sheet as shown in Appendix 3 is supplied to the contractor quoting this report and including specified values for the measured quantities. (author)

  3. 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.

  4. Sensitivity and Specificity of Clinical and Laboratory Otolith Function Tests.

    Science.gov (United States)

    Kumar, Lokesh; Thakar, Alok; Thakur, Bhaskar; Sikka, Kapil

    2017-10-01

    To evaluate clinic based and laboratory tests of otolith function for their sensitivity and specificity in demarcating unilateral compensated complete vestibular deficit from normal. Prospective cross-sectional study. Tertiary care hospital vestibular physiology laboratory. Control group-30 healthy adults, 20-45 years age; Case group-15 subjects post vestibular shwannoma excision or post-labyrinthectomy with compensated unilateral complete audio-vestibular loss. Otolith function evaluation by precise clinical testing (head tilt test-HTT; subjective visual vertical-SVV) and laboratory testing (headroll-eye counterroll-HR-ECR; vesibular evoked myogenic potentials-cVEMP). Sensitivity and specificity of clinical and laboratory tests in differentiating case and control subjects. Measurable test results were universally obtained with clinical otolith tests (SVV; HTT) but not with laboratory tests. The HR-ECR test did not indicate any definitive wave forms in 10% controls and 26% cases. cVEMP responses were absent in 10% controls.HTT test with normative cutoff at 2 degrees deviations from vertical noted as 93.33% sensitive and 100% specific. SVV test with normative cutoff at 1.3 degrees noted as 100% sensitive and 100% specific. Laboratory tests demonstrated poorer specificities owing primarily to significant unresponsiveness in normal controls. Clinical otolith function tests, if conducted with precision, demonstrate greater ability than laboratory testing in discriminating normal controls from cases with unilateral complete compensated vestibular dysfunction.

  5. Quality assurance of radiopharmaceuticals-specifications and test procedures

    International Nuclear Information System (INIS)

    Baldas, J.; Bonnyman, J.; Pojer, P.M.

    1981-08-01

    This report is a compilation of test methods used and specifications adopted for the Radiopharmaceutical Quality Assurance Test Programme conducted by the Australian Radiation Laboratory. In some cases test procedures described have been taken from various Pharmacopoeias or methods published in the literature. In other cases test methods have been developed at the ARL

  6. 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.

  7. 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)

  8. Variation of a test's sensitivity and specificity with disease prevalence

    NARCIS (Netherlands)

    Leeflang, Mariska M. G.; Rutjes, Anne W. S.; Reitsma, Johannes B.; Hooft, Lotty; Bossuyt, Patrick M. M.

    2013-01-01

    Anecdotal evidence suggests that the sensitivity and specificity of a diagnostic test may vary with disease prevalence. Our objective was to investigate the associations between disease prevalence and test sensitivity and specificity using studies of diagnostic accuracy. We used data from 23

  9. Working memory templates are maintained as feature-specific perceptual codes.

    Science.gov (United States)

    Sreenivasan, Kartik K; Sambhara, Deepak; Jha, Amishi P

    2011-07-01

    Working memory (WM) representations serve as templates that guide behavior, but the neural basis of these templates remains elusive. We tested the hypothesis that WM templates are maintained by biasing activity in sensoriperceptual neurons that code for features of items being held in memory. Neural activity was recorded using event-related potentials (ERPs) as participants viewed a series of faces and responded when a face matched a target face held in WM. Our prediction was that if activity in neurons coding for the features of the target is preferentially weighted during maintenance of the target, then ERP activity evoked by a nontarget probe face should be commensurate with the visual similarity between target and probe. Visual similarity was operationalized as the degree of overlap in visual features between target and probe. A face-sensitive ERP response was modulated by target-probe similarity. Amplitude was largest for probes that were similar to the target, and decreased monotonically as a function of decreasing target-probe similarity. These results indicate that neural activity is weighted in favor of visual features that comprise an actively held memory representation. As such, our findings support the notion that WM templates rely on neural populations involved in forming percepts of memory items.

  10. Inheritance-mode specific pathogenicity prioritization (ISPP) for human protein coding genes.

    Science.gov (United States)

    Hsu, Jacob Shujui; Kwan, Johnny S H; Pan, Zhicheng; Garcia-Barcelo, Maria-Mercè; Sham, Pak Chung; Li, Miaoxin

    2016-10-15

    Exome sequencing studies have facilitated the detection of causal genetic variants in yet-unsolved Mendelian diseases. However, the identification of disease causal genes among a list of candidates in an exome sequencing study is still not fully settled, and it is often difficult to prioritize candidate genes for follow-up studies. The inheritance mode provides crucial information for understanding Mendelian diseases, but none of the existing gene prioritization tools fully utilize this information. We examined the characteristics of Mendelian disease genes under different inheritance modes. The results suggest that Mendelian disease genes with autosomal dominant (AD) inheritance mode are more haploinsufficiency and de novo mutation sensitive, whereas those autosomal recessive (AR) genes have significantly more non-synonymous variants and regulatory transcript isoforms. In addition, the X-linked (XL) Mendelian disease genes have fewer non-synonymous and synonymous variants. As a result, we derived a new scoring system for prioritizing candidate genes for Mendelian diseases according to the inheritance mode. Our scoring system assigned to each annotated protein-coding gene (N = 18 859) three pathogenic scores according to the inheritance mode (AD, AR and XL). This inheritance mode-specific framework achieved higher accuracy (area under curve  = 0.84) in XL mode. The inheritance-mode specific pathogenicity prioritization (ISPP) outperformed other well-known methods including Haploinsufficiency, Recessive, Network centrality, Genic Intolerance, Gene Damage Index and Gene Constraint scores. This systematic study suggests that genes manifesting disease inheritance modes tend to have unique characteristics. ISPP is included in KGGSeq v1.0 (http://grass.cgs.hku.hk/limx/kggseq/), and source code is available from (https://github.com/jacobhsu35/ISPP.git). mxli@hku.hkSupplementary information: Supplementary data are available at Bioinformatics online. © The Author

  11. 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)

  12. 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

  13. Tests of the TRAC code against known analytical solutions for stratified flow

    International Nuclear Information System (INIS)

    Black, P.S.; Leslie, D.C.; Hewitt, G.F.

    1987-01-01

    The area averaged equations for gas-liquid flow are briefly summarized and related, for the specific case of stratified flow, to the shallow water equations commonly used in hydraulics. These equations are then compared to the equations used in TRAC-PF/MOD1 and are shown to differ in their treatment of the gravity head terms. A modification of the TRAC code is therefore necessary to bring it into line with established shallow water theory. The corrected form of the code was compared with a number of specific cases, each of which throws further light on the code behavior. The following areas are discussed in the paper: (1) the dam break problem; (2) Kelvin-Helmholtz instability; (3) counter-current flow; and (4) slug flow. It is concluded that detailed comparisons of the code with known analytic solutions and with a number of the more complex phenomenological experiments can give useful insights into its behavior

  14. Test specification for decant pump and winch assembly. Revision 2

    International Nuclear Information System (INIS)

    Staehr, T.W.

    1995-01-01

    This specification provides the requirements for testing of the vertical turbine decant pump including the floating suction with load sensing winch control, instrumentation and the associated PLC/PC control system. All assembly necessary for testing including piping, temporary wiring, etc., shall be performed by the Seller. All referenced figures are at the back of this document. The testing consists of performance testing, winch testing and calibration, instrumentation verification testing and run-in testing of the pump. Testing shall be done in the presence and under the direction of the Buyer in accordance with this procedure

  15. Prostate-specific antigen testing accuracy in community practice

    Directory of Open Access Journals (Sweden)

    Adams-Cameron Meg

    2002-10-01

    Full Text Available Abstract Background Most data on prostate-specific antigen (PSA testing come from urologic cohorts comprised of volunteers for screening programs. We evaluated the diagnostic accuracy of PSA testing for detecting prostate cancer in community practice. Methods PSA testing results were compared with a reference standard of prostate biopsy. Subjects were 2,620 men 40 years and older undergoing (PSA testing and biopsy from 1/1/95 through 12/31/98 in the Albuquerque, New Mexico metropolitan area. Diagnostic measures included the area under the receiver-operating characteristic curve, sensitivity, specificity, and likelihood ratios. Results Cancer was detected in 930 subjects (35%. The area under the ROC curve was 0.67 and the PSA cutpoint of 4 ng/ml had a sensitivity of 86% and a specificity of 33%. The likelihood ratio for a positive test (LR+ was 1.28 and 0.42 for a negative test (LR-. PSA testing was most sensitive (90% but least specific (27% in older men. Age-specific reference ranges improved specificity in older men (49% but decreased sensitivity (70%, with an LR+ of 1.38. Lowering the PSA cutpoint to 2 ng/ml resulted in a sensitivity of 95%, a specificity of 20%, and an LR+ of 1.19. Conclusions PSA testing had fair discriminating power for detecting prostate cancer in community practice. The PSA cutpoint of 4 ng/ml was sensitive but relatively non-specific and associated likelihood ratios only moderately revised probabilities for cancer. Using age-specific reference ranges and a PSA cutpoint below 4 ng/ml improved test specificity and sensitivity, respectively, but did not improve the overall accuracy of PSA testing.

  16. Drilling and testing specifications for the McGee well

    International Nuclear Information System (INIS)

    Patterson, J.K.

    1982-01-01

    The McGee Well is a part of the Basalt Waste Isolation Project's subsurface site selection and characterization activities. Information from the McGee Well support site hydrologic characterization and repository design. These test specifications include details for the drilling and testing of the McGee. It includes the predicted stratigraphy, the drilling requirements, description of tests to be conducted, intervals selected for hydrologic testing, and a schedule of the drilling and testing activities. 19 refs., 10 figs., 7 tabs

  17. Specific model for a gas distribution analysis in the containment at Almaraz NPP using GOTHIC computer code

    International Nuclear Information System (INIS)

    García González, M.; García Jiménez, P.; Martínez Domínguez, F.

    2016-01-01

    To carry out an analysis of the distribution of gases within the containment building at the CN Almaraz site, a simulation model with the thermohydraulic GOTHIC [1] code has been used. This has been assessed with a gas control system based on passive autocatalytic recombiners (PARs). The model is used to test the effectiveness of the control systems for gases to be used in the Almaraz Nuclear Power Plant, Uits I&II (Caceres, Spain, 1,035 MW and 1,044 MW). The model must confirm the location and number of the recombiners proposed to be installed. It is an essential function of the gas control system to avoid any formation of explosive atmospheres by reducing and limiting the concentration of combustible gases during an accident, thus maintaining the integrity of the containment. The model considers severe accident scenarios with specific conditions that produce the most onerous generation of combustible gases.

  18. 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

  19. Geometrical modification transfer between specific meshes of each coupled physical codes. Application to the Jules Horowitz research reactor experimental devices

    International Nuclear Information System (INIS)

    Duplex, B.

    2011-01-01

    The CEA develops and uses scientific software, called physical codes, in various physical disciplines to optimize installation and experimentation costs. During a study, several physical phenomena interact, so a code coupling and some data exchanges between different physical codes are required. Each physical code computes on a particular geometry, usually represented by a mesh composed of thousands to millions of elements. This PhD Thesis focuses on the geometrical modification transfer between specific meshes of each coupled physical code. First, it presents a physical code coupling method where deformations are computed by one of these codes. Next, it discusses the establishment of a model, common to different physical codes, grouping all the shared data. Finally, it covers the deformation transfers between meshes of the same geometry or adjacent geometries. Geometrical modifications are discrete data because they are based on a mesh. In order to permit every code to access deformations and to transfer them, a continuous representation is computed. Two functions are developed, one with a global support, and the other with a local support. Both functions combine a simplification method and a radial basis function network. A whole use case is dedicated to the Jules Horowitz reactor. The effect of differential dilatations on experimental device cooling is studied. (author) [fr

  20. Templates for Cross-Cultural and Culturally Specific Usability Testing

    DEFF Research Database (Denmark)

    Clemmensen, Torkil

    2011-01-01

    The cultural diversity of users of technology challenges our methods for usability testing. This article suggests templates for cross-culturally and culturally specific usability testing, based on studies of usability testing in companies in Mumbai, Beijing, and Copenhagen. Study 1 was a cross...... tests. The result was the construction of templates for usability testing. The culturally specific templates were in Mumbai “user-centered evaluation,” Copenhagen “client-centered evaluation,” and Beijing “evaluator-centered evaluation.” The findings are compared with related research...

  1. 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.

  2. 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

  3. 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)

  4. 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'

  5. 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)

  6. Formal Specification Based Automatic Test Generation for Embedded Network Systems

    Directory of Open Access Journals (Sweden)

    Eun Hye Choi

    2014-01-01

    Full Text Available Embedded systems have become increasingly connected and communicate with each other, forming large-scaled and complicated network systems. To make their design and testing more reliable and robust, this paper proposes a formal specification language called SENS and a SENS-based automatic test generation tool called TGSENS. Our approach is summarized as follows: (1 A user describes requirements of target embedded network systems by logical property-based constraints using SENS. (2 Given SENS specifications, test cases are automatically generated using a SAT-based solver. Filtering mechanisms to select efficient test cases are also available in our tool. (3 In addition, given a testing goal by the user, test sequences are automatically extracted from exhaustive test cases. We’ve implemented our approach and conducted several experiments on practical case studies. Through the experiments, we confirmed the efficiency of our approach in design and test generation of real embedded air-conditioning network systems.

  7. Evidence for gene-specific rather than transcription rate-dependent histone H3 exchange in yeast coding regions.

    Science.gov (United States)

    Gat-Viks, Irit; Vingron, Martin

    2009-02-01

    In eukaryotic organisms, histones are dynamically exchanged independently of DNA replication. Recent reports show that different coding regions differ in their amount of replication-independent histone H3 exchange. The current paradigm is that this histone exchange variability among coding regions is a consequence of transcription rate. Here we put forward the idea that this variability might be also modulated in a gene-specific manner independently of transcription rate. To that end, we study transcription rate-independent replication-independent coding region histone H3 exchange. We term such events relative exchange. Our genome-wide analysis shows conclusively that in yeast, relative exchange is a novel consistent feature of coding regions. Outside of replication, each coding region has a characteristic pattern of histone H3 exchange that is either higher or lower than what was expected by its RNAPII transcription rate alone. Histone H3 exchange in coding regions might be a way to add or remove certain histone modifications that are important for transcription elongation. Therefore, our results that gene-specific coding region histone H3 exchange is decoupled from transcription rate might hint at a new epigenetic mechanism of transcription regulation.

  8. Test specification for decant pump and winch assembly

    International Nuclear Information System (INIS)

    Staehr, T.W.

    1994-01-01

    This specification provides the requirements for testing of the vertical turbine decant pump including the floating suction arm with load sensing winch control, instrumentation and the associated PLC/PC control system

  9. Quality assurance of radiopharmaceuticals - specifications and test procedures

    International Nuclear Information System (INIS)

    Baldas, J.; Bonnyman, J.; Colmanet, S.F.; Ivanov, Z.; Lauder, R.A.

    1990-10-01

    The authors report on a Radiopharmaceutical Quality Assurance Test Programme carried out by the Australian Radiation Laboratory in which radiopharmaceuticals used in nuclear medicine in Australia are tested for compliance with specifications. Where the radiopharmaceutical is the subject of a monograph in the British Pharmacopoeia or the European Pharmacopoeia, then the specifications given in the Pharmacopoeia are adopted. In other cases the specifications given have been adopted by this Laboratory and have no legal status. In some cases test procedures described have been taken from various Pharmacopoeias or methods published in the literature. In other cases test methods described have been developed at this Laboratory. It should be noted that, unless stated otherwise, specifications listed apply at all times up until product expire

  10. 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

  11. 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.

  12. 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.

  13. Sensitivity and specificity of neuropsychological tests for dementia

    African Journals Online (AJOL)

    specificity of a battery of neuropsychological tests in a sample of elderly persons living in a ... estimate of 20% prevalence for dementia in residential homes ... demographic variables, and mean neuro- psychological .... on optimum balance between sensitivity and specificity (Fig. 1). ..... The lack of stratification of the sample.

  14. 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

  15. Cerebellum-specific and age-dependent expression of an endogenous retrovirus with intact coding potential

    Directory of Open Access Journals (Sweden)

    Itoh Takayuki

    2011-10-01

    Full Text Available Abstract Background Endogenous retroviruses (ERVs, including murine leukemia virus (MuLV type-ERVs (MuLV-ERVs, are presumed to occupy ~10% of the mouse genome. In this study, following the identification of a full-length MuLV-ERV by in silico survey of the C57BL/6J mouse genome, its distribution in different mouse strains and expression characteristics were investigated. Results Application of a set of ERV mining protocols identified a MuLV-ERV locus with full coding potential on chromosome 8 (named ERVmch8. It appears that ERVmch8 shares the same genomic locus with a replication-incompetent MuLV-ERV, called Emv2; however, it was not confirmed due to a lack of relevant annotation and Emv2 sequence information. The ERVmch8 sequence was more prevalent in laboratory strains compared to wild-derived strains. Among 16 different tissues of ~12 week-old female C57BL/6J mice, brain homogenate was the only tissue with evident expression of ERVmch8. Further ERVmch8 expression analysis in six different brain compartments and four peripheral neuronal tissues of C57BL/6J mice revealed no significant expression except for the cerebellum in which the ERVmch8 locus' low methylation status was unique compared to the other brain compartments. The ERVmch8 locus was found to be surrounded by genes associated with neuronal development and/or inflammation. Interestingly, cerebellum-specific ERVmch8 expression was age-dependent with almost no expression at 2 weeks and a plateau at 6 weeks. Conclusions The ecotropic ERVmch8 locus on the C57BL/6J mouse genome was relatively undermethylated in the cerebellum, and its expression was cerebellum-specific and age-dependent.

  16. 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.

  17. 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

  18. 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

  19. Low specificity of 2 tetanus rapid tests in Cambodia.

    Science.gov (United States)

    Schlumberger, M; Yvonnet, B; Lesage, G; Tep, B

    2015-01-01

    Rapid testing for tetanus on serum or blood allows for an immediate evaluation of individual protection against tetanus in developed countries, using a "single step" immunochromatographic technique using tetanus toxoid. The specificity of these tests, compared to the reference method for tetanus, mouse serum neutralization testing, has however never been assessed in these countries, due to the difficulty to perform serum neutralization titration in mice, because of animal testing bioethical regulations. A collection of sera from adult volunteers in Cambodia, living in rural environment, was tested for tetanus antibodies by ELISA in France, and by mouse serum neutralization in Vietnam. This allowed estimating the sensitivity and specificity of 2 rapid tetanus tests, available on the market: TQS™ and Tetanotop™. The sensitivity of these tests was adequate, compared to mice serum neutralization test, for a test threshold of 0.01 IU/mL, (100% for TQS™, 91% for Tetanotop™), but their specificity was very low (1% for TQS™ and 13% for Tetanotop™). The results prove that these rapid tests for the assessment of individual protection against tetanus should not be used in the adult rural Cambodian population. Copyright © 2015 Elsevier Masson SAS. All rights reserved.

  20. Comparison of the calculations of the stability properties of a specific stellarator equilibrium with different MHD stability codes

    International Nuclear Information System (INIS)

    Nakamura, Y.; Matsumoto, T.; Wakatani, M.; Ichiguchi, K.; Garcia, L.; Carreras, B.A.

    1995-04-01

    A particular configuration of the LHD stellarator with an unusually flat pressure profile has been chosen to be a test case for comparison of the MHD stability property predictions of different three-dimensional and averaged codes for the purpose of code comparison and validation. In particular, two relatively localized instabilities, the fastest growing modes with toroidal mode number n = 2 and n = 3 were studied using several different codes, with the good agreement that has been found providing justification for the use of any of them for equilibria of the type considered

  1. 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)

  2. OSSMETER D3.4 – Language-Specific Source Code Quality Analysis

    NARCIS (Netherlands)

    J.J. Vinju (Jurgen); A. Shahi (Ashim); H.J.S. Basten (Bas)

    2014-01-01

    htmlabstractThis deliverable is part of WP3: Source Code Quality and Activity Analysis. It provides descriptions and prototypes of the tools that are needed for source code quality analysis in open source software projects. It builds upon the results of: • Deliverable 3.1 where infra-structure and

  3. Lightweight Detection of Android-specific Code Smells : The aDoctor Project

    NARCIS (Netherlands)

    Palomba, F.; Di Nucci, D.; Panichella, A.; Zaidman, A.E.; De Lucia, Andrea; Pinzger, Martin; Bavota, Gabriele; Marcus, Andrian

    2017-01-01

    Code smells are symptoms of poor design solutions applied by programmers during the development of software systems. While the research community devoted a lot of effort to studying and devising approaches for detecting the traditional code smells defined by Fowler, little knowledge and support

  4. Technical Specifications of Structural Health Monitoring for Highway Bridges: New Chinese Structural Health Monitoring Code

    Directory of Open Access Journals (Sweden)

    Fernando Moreu

    2018-03-01

    Full Text Available Governments and professional groups related to civil engineering write and publish standards and codes to protect the safety of critical infrastructure. In recent decades, countries have developed codes and standards for structural health monitoring (SHM. During this same period, rapid growth in the Chinese economy has led to massive development of civil engineering infrastructure design and construction projects. In 2016, the Ministry of Transportation of the People’s Republic of China published a new design code for SHM systems for large highway bridges. This document is the first technical SHM code by a national government that enforces sensor installation on highway bridges. This paper summarizes the existing international technical SHM codes for various countries and compares them with the new SHM code required by the Chinese Ministry of Transportation. This paper outlines the contents of the new Chinese SHM code and explains its relevance for the safety and management of large bridges in China, introducing key definitions of the Chinese–United States SHM vocabulary and their technical significance. Finally, this paper discusses the implications for the design and implementation of a future SHM codes, with suggestions for similar efforts in United States and other countries.

  5. 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

  6. 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.

  7. Sensitivity and specificity of the nickel spot (dimethylglyoxime) test.

    Science.gov (United States)

    Thyssen, Jacob P; Skare, Lizbet; Lundgren, Lennart; Menné, Torkil; Johansen, Jeanne D; Maibach, Howard I; Lidén, Carola

    2010-05-01

    The accuracy of the dimethylglyoxime (DMG) nickel spot test has been questioned because of false negative and positive test reactions. The EN 1811, a European standard reference method developed by the European Committee for Standardization (CEN), is fine-tuned to estimate nickel release around the limit value of the EU Nickel Directive from products intended to come into direct and prolonged skin contact. Because assessments according to EN 1811 are expensive to perform, time consuming, and may destruct the test item, it should be of great value to know the accuracy of the DMG screening test. To evaluate the sensitivity and specificity of the DMG test. DMG spot testing, chemical analysis according to the EN 1811 reference method, and X-ray fluorescence spectroscopy (XRF) were performed concomitantly on 96 metallic components from earrings recently purchased in San Francisco. The sensitivity of the DMG test was 59.3% and the specificity was 97.5% based on DMG-test results and nickel release concentrations determined by the EN 1811 reference method. The DMG test has a high specificity but a modest sensitivity. It may serve well for screening purposes. Past exposure studies may have underestimated nickel release from consumer items.

  8. 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)

  9. Testing Moral Foundation Theory: Are Specific Moral Emotions Elicited by Specific Moral Transgressions?

    Science.gov (United States)

    Landmann, Helen; Hess, Ursula

    2018-01-01

    Moral foundation theory posits that specific moral transgressions elicit specific moral emotions. To test this claim, participants (N = 195) were asked to rate their emotions in response to moral violation vignettes. We found that compassion and disgust were associated with care and purity respectively as predicted by moral foundation theory.…

  10. Divergent evolutionary rates in vertebrate and mammalian specific conserved non-coding elements (CNEs) in echolocating mammals.

    Science.gov (United States)

    Davies, Kalina T J; Tsagkogeorga, Georgia; Rossiter, Stephen J

    2014-12-19

    The majority of DNA contained within vertebrate genomes is non-coding, with a certain proportion of this thought to play regulatory roles during development. Conserved Non-coding Elements (CNEs) are an abundant group of putative regulatory sequences that are highly conserved across divergent groups and thus assumed to be under strong selective constraint. Many CNEs may contain regulatory factor binding sites, and their frequent spatial association with key developmental genes - such as those regulating sensory system development - suggests crucial roles in regulating gene expression and cellular patterning. Yet surprisingly little is known about the molecular evolution of CNEs across diverse mammalian taxa or their role in specific phenotypic adaptations. We examined 3,110 vertebrate-specific and ~82,000 mammalian-specific CNEs across 19 and 9 mammalian orders respectively, and tested for changes in the rate of evolution of CNEs located in the proximity of genes underlying the development or functioning of auditory systems. As we focused on CNEs putatively associated with genes underlying the development/functioning of auditory systems, we incorporated echolocating taxa in our dataset because of their highly specialised and derived auditory systems. Phylogenetic reconstructions of concatenated CNEs broadly recovered accepted mammal relationships despite high levels of sequence conservation. We found that CNE substitution rates were highest in rodents and lowest in primates, consistent with previous findings. Comparisons of CNE substitution rates from several genomic regions containing genes linked to auditory system development and hearing revealed differences between echolocating and non-echolocating taxa. Wider taxonomic sampling of four CNEs associated with the homeobox genes Hmx2 and Hmx3 - which are required for inner ear development - revealed family-wise variation across diverse bat species. Specifically within one family of echolocating bats that utilise

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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.

  17. 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)

  18. Primate-specific spliced PMCHL RNAs are non-protein coding in human and macaque tissues

    Directory of Open Access Journals (Sweden)

    Delerue-Audegond Audrey

    2008-12-01

    Full Text Available Abstract Background Brain-expressed genes that were created in primate lineage represent obvious candidates to investigate molecular mechanisms that contributed to neural reorganization and emergence of new behavioural functions in Homo sapiens. PMCHL1 arose from retroposition of a pro-melanin-concentrating hormone (PMCH antisense mRNA on the ancestral human chromosome 5p14 when platyrrhines and catarrhines diverged. Mutations before divergence of hylobatidae led to creation of new exons and finally PMCHL1 duplicated in an ancestor of hominids to generate PMCHL2 at the human chromosome 5q13. A complex pattern of spliced and unspliced PMCHL RNAs were found in human brain and testis. Results Several novel spliced PMCHL transcripts have been characterized in human testis and fetal brain, identifying an additional exon and novel splice sites. Sequencing of PMCHL genes in several non-human primates allowed to carry out phylogenetic analyses revealing that the initial retroposition event took place within an intron of the brain cadherin (CDH12 gene, soon after platyrrhine/catarrhine divergence, i.e. 30–35 Mya, and was concomitant with the insertion of an AluSg element. Sequence analysis of the spliced PMCHL transcripts identified only short ORFs of less than 300 bp, with low (VMCH-p8 and protein variants or no evolutionary conservation. Western blot analyses of human and macaque tissues expressing PMCHL RNA failed to reveal any protein corresponding to VMCH-p8 and protein variants encoded by spliced transcripts. Conclusion Our present results improve our knowledge of the gene structure and the evolutionary history of the primate-specific chimeric PMCHL genes. These genes produce multiple spliced transcripts, bearing short, non-conserved and apparently non-translated ORFs that may function as mRNA-like non-coding RNAs.

  19. Novel speed test for evaluation of badminton specific movements

    DEFF Research Database (Denmark)

    Madsen, Christian Møller; Karlsen, Anders; Nybo, Lars

    2015-01-01

    In this study we developed a novel badminton speed test (BST). The test was designed to mimic match play. The test starts in the center of the court and consists of five maximal actions to sensors located in each of the four corners of the court. The 20 actions are performed in randomized order...... as dictated by computer screen shots displayed one second following completion of the previous action. We assessed day-to-day variation in elite players and specificity of the test was evaluated by comparing 30 meter sprint performance and time to complete the BST in 20 elite, 21 skilled players and 20 age...

  20. 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

  1. 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

  2. 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.

  3. Test and evaluation document for DOT Specification 7A Type A Packaging. Revision 3

    International Nuclear Information System (INIS)

    1996-01-01

    The US Department of Energy (DOE) has been conducting, through several of its operating contractors, an evaluation and testing program to qualify Type A radioactive material packagings per US Department of Transportation (DOT) Specification 7A (DOT-7A) of the Code of Federal Regulations (CFR), Title 49, Part 178 (49 CFR 178). The program is currently administered by the DOE, Office of Facility Safety Analysis, DOE/EH-32, at DOE-Headquarters (DOE-HQ) in Germantown, Maryland. This document summarizes the evaluation and testing performed for all of the packagings successfully qualified in this program

  4. Test and evaluation document for DOT Specification 7A Type A Packaging. Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-01-30

    The US Department of Energy (DOE) has been conducting, through several of its operating contractors, an evaluation and testing program to qualify Type A radioactive material packagings per US Department of Transportation (DOT) Specification 7A (DOT-7A) of the Code of Federal Regulations (CFR), Title 49, Part 178 (49 CFR 178). The program is currently administered by the DOE, Office of Facility Safety Analysis, DOE/EH-32, at DOE-Headquarters (DOE-HQ) in Germantown, Maryland. This document summarizes the evaluation and testing performed for all of the packagings successfully qualified in this program.

  5. 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)

  6. 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)

  7. 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)

  8. 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)

  9. Variation of a test's sensitivity and specificity with disease prevalence.

    Science.gov (United States)

    Leeflang, Mariska M G; Rutjes, Anne W S; Reitsma, Johannes B; Hooft, Lotty; Bossuyt, Patrick M M

    2013-08-06

    Anecdotal evidence suggests that the sensitivity and specificity of a diagnostic test may vary with disease prevalence. Our objective was to investigate the associations between disease prevalence and test sensitivity and specificity using studies of diagnostic accuracy. We used data from 23 meta-analyses, each of which included 10-39 studies (416 total). The median prevalence per review ranged from 1% to 77%. We evaluated the effects of prevalence on sensitivity and specificity using a bivariate random-effects model for each meta-analysis, with prevalence as a covariate. We estimated the overall effect of prevalence by pooling the effects using the inverse variance method. Within a given review, a change in prevalence from the lowest to highest value resulted in a corresponding change in sensitivity or specificity from 0 to 40 percentage points. This effect was statistically significant (p disease prevalence; there was no such systematic effect for sensitivity. The sensitivity and specificity of a test often vary with disease prevalence; this effect is likely to be the result of mechanisms, such as patient spectrum, that affect prevalence, sensitivity and specificity. Because it may be difficult to identify such mechanisms, clinicians should use prevalence as a guide when selecting studies that most closely match their situation.

  10. Specifications for a two-dimensional multi-group scattering code: ALCI; Specification d'un code de diffusion multigroupe a deux dimensions: ALCI

    Energy Technology Data Exchange (ETDEWEB)

    Bayard, J P; Guillou, A; Lago, B; Bureau du Colombier, M J; Guillou, G; Vasseur, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, R{theta}. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [French] Ce rapport decrit les specifications du programme ALCI. Ce programme resout le systeme d'equations aux differences approchant le probleme homogene de la diffusion neutronique multigroupe, a deux dimensions d'espace, dans les trois geometries XY, RZ, R{theta}. Il permet des calculs de criticalite geometrique et de composition et calcule sur demande le probleme adjoint. Le nombre maximum de points traites est de 6000. Le nombre maximum de groupes permis est de 12. Les iterations interieure sont traitees par la methode des directions alternees. Les iterations exterieures sont accelerees par la methode d'extrapolation de Tchebychev. (auteurs)

  11. Subsurface barrier demonstration test strategy and performance specification

    International Nuclear Information System (INIS)

    Treat, R.L.; Cruse, J.M.

    1994-05-01

    This document was developed to help specify a major demonstration test project of subsurface barrier systems supporting the Tank Waste Remediation System (TWRS) Program. The document focuses discussion on requirements applicable to demonstration of three subsurface barrier concepts: (1) Injected Material, (2) Cryogenic, and (3) Desiccant. Detailed requirements are provided for initial qualification of a technology proposal followed by the pre-demonstration and demonstration test requirements and specifications. Each requirement and specification is accompanied by a discussion of the rationale for it. The document also includes information on the Hanford Site tank farms and related data; the related and currently active technology development projects within the DOE's EM-50 Program; and the overall demonstration test strategy. Procurement activities and other preparations for actual demonstration testing are on hold until a decision is made regarding further development of subsurface barriers. Accordingly, this document is being issued for information only

  12. Specification and acceptance testing of radiotherapy treatment planning systems

    International Nuclear Information System (INIS)

    2007-04-01

    Quality assurance (QA) in the radiation therapy treatment planning process is essential to ensure accurate dose delivery to the patient and to minimize the possibility of accidental exposure. The computerized radiotherapy treatment planning systems (RTPSs) are now widely available in industrialized and developing countries and it is of special importance to support hospitals in Member States in developing procedures for acceptance testing, commissioning and QA of their RTPSs. Responding to these needs, a group of experts developed an IAEA publication with such recommendations, which was published in 2004 as IAEA Technical Reports Series No. 430. This report provides a general framework and describes a large number of tests and procedures that should be considered by the users of new RTPSs. However, small hospitals with limited resources or large hospitals with high patient load and limited staff are not always able to perform complete characterization, validation and software testing of algorithms used in RTPSs. Therefore, the IAEA proposed more specific guidelines that provide a step-by-step recommendation for users at hospitals or cancer centres how to implement acceptance and commissioning procedures for newly purchased RTPSs. The current publication was developed in the framework of the Coordinated Research Project on Development of Procedures for Quality Assurance for Dosimetry Calculations in Radiotherapy and uses the International Electrotechnical Commission (IEC) standard IEC 62083, Requirements for the Safety of Radiotherapy Treatment Planning Systems as its basis. The report addresses the procedures for specification and acceptance testing of RTPSs to be used by both manufacturers and users at the hospitals. Recommendations are provided for specific tests to be performed at the manufacturing facility known as type tests, and for acceptance tests to be performed at the hospital known as site tests. The purpose of acceptance testing is to demonstrate to the

  13. 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

  14. 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.

  15. Specificity Protein (Sp) Transcription Factors and Metformin Regulate Expression of the Long Non-coding RNA HULC

    Science.gov (United States)

    There is evidence that specificity protein 1 (Sp1) transcription factor (TF) regulates expression of long non-coding RNAs (lncRNAs) in hepatocellular carcinoma (HCC) cells. RNA interference (RNAi) studies showed that among several lncRNAs expressed in HepG2, SNU-449 and SK-Hep-1...

  16. 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.

  17. 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.

  18. PBF/LOFT Lead Rod Test Program experiment operating specification

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.

    1978-11-01

    The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory. Understanding the behavior of light-water reactors (LWR) under loss-of-coolant conditions is a major objective of the NRC Reactor Safety Research Program. The Loss of Fluid Test (LOFT) facility is the major testing facility to evaluate the systems response of an LWR over a wide range of Loss of Coolant Experment (LOCE) conditions. As such, the LOFT core is intended to be used for sequential LOCE tests provided no significant fuel rod failures occur. The PFB/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-2 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel over a wide range of power. Thus, a relatively accurate assessment of the state of the LOFT core after the completion of each subtest and the anticipated effect of the next test can be obtained by utilizing a combination of LLR test data and analytical predictions. Specifications for the test program are presented

  19. 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.

  20. Procurement specification high vacuum test chamber and pumping system

    International Nuclear Information System (INIS)

    1976-01-01

    The specification establishes requirements for a high-vacuum test chamber, associated vacuum pumps, valves, controls, and instrumentation that shall be designed and fabricated for use as a test chamber for testing a closed loop Brayton Isotope Power System (BIPS) Ground Demonstration System (GDS). The vacuum system shall include all instrumentation required for pressure measurement and control of the vacuum pumping system. A general outline of the BIPS-GDS in the vacuum chamber and the preliminary piping and instrumentation interface to the vacuum chamber are shown

  1. 40 CFR 798.5195 - Mouse biochemical specific locus test.

    Science.gov (United States)

    2010-07-01

    ...-induced variants are bred to determine the genetic nature of the change. (f) Data and reports—(1... SUBSTANCES CONTROL ACT (CONTINUED) HEALTH EFFECTS TESTING GUIDELINES Genetic Toxicity § 798.5195 Mouse...) A biochemical specific locus mutation is a genetic change resulting from a DNA lesion causing...

  2. Satellite III non-coding RNAs show distinct and stress-specific patterns of induction

    International Nuclear Information System (INIS)

    Sengupta, Sonali; Parihar, Rashmi; Ganesh, Subramaniam

    2009-01-01

    The heat shock response in human cells is associated with the transcription of satellite III repeats (SatIII) located in the 9q12 locus. Upon induction, the SatIII transcripts remain associated with the locus and recruit several transcription and splicing factors to form the nuclear stress bodies (nSBs). The nSBs are thought to modulate epigenetic changes during the heat shock response. We demonstrate here that the nSBs are induced by a variety of stressors and show stress-specific patterns of induction. While the transcription factor HSF1 is required for the induction of SatIII locus by the stressors tested, its specific role in the transcriptional process appears to be stress dependent. Our results suggest the existence of multiple transcriptional loci for the SatIII transcripts and that their activation might depend upon the type of stressors. Thus, induction of SatIII transcripts appears to be a generic response to a variety of stress conditions.

  3. 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

  4. Standard Specification for Solar Simulation for Terrestrial Photovoltaic Testing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification provides means for classifying solar simulators intended for indoor testing of photovoltaic devices (solar cells or modules), according to their spectral match to a reference spectral irradiance, non-uniformity of spatial irradiance, and temporal instability of irradiance. 1.2 Testing of photovoltaic devices may require the use of solar simulators. Test Methods that require specific classification of simulators as defined in this specification include Test Methods E948, E1036, and E1362. 1.3 This standard is applicable to both pulsed and steady state simulators and includes recommended test requirements used for classifying such simulators. 1.4 A solar simulator usually consists of three major components: (1) light source(s) and associated power supply; (2) any optics and filters required to modify the output beam to meet the classification requirements in Section 4; and (3) the necessary controls to operate the simulator, adjust irradiance, etc. 1.5 A light source that does not mee...

  5. Critical evaluation of a badminton-specific endurance test.

    Science.gov (United States)

    Fuchs, Michael; Faude, Oliver; Wegmann, Melissa; Meyer, Tim

    2014-03-01

    To overcome the limitations of traditional 1-dimensional fitness tests in analyzing physiological properties of badminton players, a badminton-specific endurance test (BST) was created. This study aimed at analyzing the influence of various fitness dimensions on BST performance. 18 internationally competing male German badminton players (22.4 ± 3.2 y, 79.2 ± 7.7 kg, 1.84 ± 0.06 m, world-ranking position [WRP] 21-501) completed a straight-sprint test, a change-of-direction speed test, various jump tests (countermovement jump, drop jump, standing long jump), a multistage running test (MST), and the BST. During this on-court field test players have to respond to a computerized sign indicating direction and speed of badminton-specific movements by moving into the corresponding corners. Significant correlations were found between performance in MST and BST (individual anaerobic threshold [IAT], r = .63, P = .005; maximum velocity [Vmax], r = .60, P = .009). A negative correlation (r = -.59, P = .014) was observed between IAT in BST and drop-jump contact time. No further associations between performance indices could be detected. Apart from a small portion explained by MST results (IAT, R2 = .40; Vmax, R2 = .36), the majority of BST performance cannot be explained by the determined physiological correlates. Moreover, it was impossible to predict the WRP of a player on the basis of BST results (r = -.15, P = .55). Neither discipline-specific performance nor basic physiological properties were appropriately reflected by a BST in elite badminton players. This does not substantiate its validity for regular use as a testing tool. However, it may be useful for monitoring on-court training sessions.

  6. 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.

  7. Formal Functional Test Designs: Bridging the Gap Between Test Requirements and Test Specifications

    Science.gov (United States)

    Hops, Jonathan

    1993-01-01

    This presentation describes the testing life cycle, the purpose of the test design phase, and test design methods and gives an example application. Also included is a description of Test Representation Language (TRL), a summary of the language, and an example of an application of TRL. A sample test requirement and sample test design are included.

  8. 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)

  9. Rascal: A domain specific language for source code analysis and manipulation

    NARCIS (Netherlands)

    P. Klint (Paul); T. van der Storm (Tijs); J.J. Vinju (Jurgen); A. Walenstein; S. Schuppe

    2009-01-01

    htmlabstractMany automated software engineering tools require tight integration of techniques for source code analysis and manipulation. State-of-the-art tools exist for both, but the domains have remained notoriously separate because different computational paradigms fit each domain best. This

  10. RASCAL : a domain specific language for source code analysis and manipulationa

    NARCIS (Netherlands)

    Klint, P.; Storm, van der T.; Vinju, J.J.

    2009-01-01

    Many automated software engineering tools require tight integration of techniques for source code analysis and manipulation. State-of-the-art tools exist for both, but the domains have remained notoriously separate because different computational paradigms fit each domain best. This impedance

  11. Specific structural probing of plasmid-coded ribosomal RNAs from Escherichia coli

    DEFF Research Database (Denmark)

    Aagaard, C; Rosendahl, G; Dam, M

    1991-01-01

    The preferred method for construction and in vivo expression of mutagenised Escherichia coli ribosomal RNAs (rRNAs) is via high copy number plasmids. Transcription of wild-type rRNA from the seven chromosomal rrn operons in strains harbouring plasmid-coded mutant rRNAs leads to a heterogeneous...

  12. Novel speed test for evaluation of badminton-specific movements.

    Science.gov (United States)

    Madsen, Christian M; Karlsen, Anders; Nybo, Lars

    2015-05-01

    In this study, we developed a novel badminton-specific speed test (BST). The test was designed to mimic match play. The test starts in the center of the court and consists of 5 maximal actions to sensors located in each of the 4 corners of the court. The 20 actions are performed in randomized order as dictated by computer screen shots displayed 1 second after completion of the previous action. We assessed day-to-day variation in elite players, and specificity of the test was evaluated by comparing 30-m sprint performance and time to complete the BST in 20 elite players, 21 skilled players, and 20 age-matched physical active subjects (non-badminton players). Sprint performance was similar across groups, whereas the elite players were significantly (p ≤ 0.05) faster in the BST (total test time: 32.3 ± 1.1 seconds; average: 1.6 seconds per action) than the skilled (34.1 ± 2.0 seconds) and non-badminton players (35.7 ± 1.7 seconds). Day-to-day coefficient of variation (CV) of the BST was 0.7% for the elite players, whereas CV for repeated tests on the same day was 1.7% for elite, 2.6% for skilled, and 2.5% for non-badminton players. On this basis, we suggest that the BST may be valuable for evaluation of short-term maximal movement speed in badminton players. Thus, the BST seems to be sport specific, as it may discriminate between groups (elite, less trained players, and non-badminton players) with similar sprinting performance, and the low test-retest variation may allow for using the BST to evaluate longitudinal changes, for example, training effects or seasonal variations.

  13. 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

  14. Specification and testing of optics for LIS system

    International Nuclear Information System (INIS)

    Singh, Sunita; Sridhar, G.; Rawat, V.S.; Gantayet, L.M.

    2005-01-01

    Optical component specification for the high average power lasers and laser beam transport system used in the laser isotope separation demonstration facility must address demanding system performance requirements. In a typical demonstration facility a few thousand of commercial and custom optical components are required. The optical system is expected to perform at a high level of optical efficiency and reliability. Evaluation and testing of optical components used in LIS plant is critical for qualification of suppliers and assurance of performance in the actual process. The stringent specifications require specialized test equipment and techniques, which are not routine. Careful planning with the optics manufacturer, detailed quality assurance plan, comprehensive procedures for testing and evaluation, and a plan for corrective action are required. The specifications are given on material characteristics, surface quality and flatness, reflectance or transmittance and high average power laser damage. Our approach to specifying, testing the performance characteristics and assuring quality of optical components required for the technology demonstration of laser based isotopic clean-up of 233 U project is presented. (author)

  15. Biomechanical assessment of dynamic balance: Specificity of different balance tests.

    Science.gov (United States)

    Ringhof, Steffen; Stein, Thorsten

    2018-04-01

    Dynamic balance is vitally important for most sports and activities of daily living, so the assessment of dynamic stability has become an important issue. In consequence, a large number of balance tests have been developed. However, it is not yet known whether these tests (i) measure the same construct and (ii) can differentiate between athletes with different balance expertise. We therefore studied three common dynamic balance tests: one-leg jump landings, Posturomed perturbations and simulated forward falls. Participants were 24 healthy young females in regular training in either gymnastics (n = 12) or swimming (n = 12). In each of the tests, the participants were instructed to recover balance as quickly as possible. Dynamic stability was computed by time to stabilization and margin of stability, deduced from force plates and motion capture respectively. Pearson's correlations between the dynamic balance tests found no significant associations between the respective dynamic stability measures. Furthermore, independent t-tests indicated that only jump landings could properly distinguish between both groups of athletes. In essence, the different dynamic balance tests applied did not measure the same construct but rather task-specific skills, each of which depends on multifactorial internal and external constraints. Our study therefore contradicts the traditional view of considering balance as a general ability, and reinforces that dynamic balance measures are not interchangeable. This highlights the importance of selecting appropriate balance tests. Copyright © 2018 Elsevier B.V. All rights reserved.

  16. Specification and Test of Real-Time Systems

    DEFF Research Database (Denmark)

    Nielsen, Brian

    of the system, and a set of constraint patterns which describes and enforces the timing and synchronization constraints among components. We propose new techniques for automated black box conformance testing of real-time systems against densely timed speci cations. A test generator tool examines a specification......Distributed real-time computer based systems are very complex and intrinsically difficult to specify and implement correctly; in part this is caused by the overwhelming number of possible interactions between system components, but especially by a lack of adequate methods and tools to deal...... of the desired system behavior and generates the necessary test cases. A main problem is to construct a reasonably small test suite that can be executed within allotted resources, while having a high likelihood of detecting unknown errors. Our goal has been to treat the time dimension of this problem thoroughly...

  17. 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)

  18. 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)

  19. 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

  20. Specificity and sensitivity assessment of selected nasal provocation testing techniques

    Directory of Open Access Journals (Sweden)

    Edyta Krzych-Fałta

    2016-12-01

    Full Text Available Introduction: Nasal provocation testing involves an allergen-specific local reaction of the nasal mucosa to the administered allergen. Aim: To determine the most objective nasal occlusion assessment technique that could be used in nasal provocation testing. Material and methods : A total of 60 subjects, including 30 patients diagnosed with allergy to common environmental allergens and 30 healthy subjects were enrolled into the study. The method used in the study was a nasal provocation test with an allergen, with a standard dose of a control solution and an allergen (5,000 SBU/ml administered using a calibrated atomizer into both nostrils at room temperature. Early-phase nasal mucosa response in the early phase of the allergic reaction was assessed via acoustic rhinometry, optical rhinometry, nitric oxide in nasal air, and tryptase levels in the nasal lavage fluid. Results : In estimating the homogeneity of the average values, the Levene’s test was used and receiver operating characteristic curves were plotted for all the methods used for assessing the nasal provocation test with an allergen. Statistically significant results were defined for p < 0.05. Of all the objective assessment techniques, the most sensitive and characteristic ones were the optical rhinometry techniques (specificity = 1, sensitivity = 1, AUC = 1, PPV = 1, NPV = 1. Conclusions : The techniques used showed significant differences between the group of patients with allergic rhinitis and the control group. Of all the objective assessment techniques, those most sensitive and characteristic were the optical rhinometry.

  1. 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.)

  2. 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.

  3. Test specifications for the waste information and control system

    International Nuclear Information System (INIS)

    Flynn, D.F.

    1994-01-01

    This document describes the test specifications for the testing of the WICS system. The Westinghouse Hanford Company (WHC) Hazardous Material Control Group (HMC) of the 222-S Laboratory has requested the development of a system to help resolve many of the difficulties associated with tracking and data collection of containers and drums of waste. This system has been identified as Waste Information and Control System (WICS). The request for developing and implementing WICS has been made to the Automation and Simulation Engineering Group (ASE)

  4. Structured flowcharts for control logic specification in the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Nielson, C.W.; Claborn, G.W.

    1983-01-01

    The Tritium Systems Test Assembly (TSTA) contains several subsystems employing sophisticated chemical and physical processes to purify, transport, and capture the isotopes of hydrogen. The ultimate responsibility for the correct and safe operation of these subsystems lies with their designers. However, the logic is implemented in a computer system with program control. A means to insure unambiguous specification of the control logic in a form understandable to both the non-programming designers and the software staff was required. The computer programs are written in RATFOR, a language providing clear control structures and powerful symbol definition facilities. However, the actual code was considered unsatisfactory as a means of primary specification by the non-programming designers. On the other hand, simple English language descriptions of the desired behavior were not precise enough to insure correctness. Experimentation with traditional flowcharts proved that they were more difficult to follow than the RATFOR code. On the other hand, the use of structured flowcharts derived from those introduced by Nassi and Shneidermanl have proven to be very powerful. Using simple geometric forms for the basic control structures such as loops and conditional tests, and by using expansion rather than connection as the means of reducing any flowchart to a single page, a specification that is both understandable and precise has been obtained. A computer code automates the production and modification of these flowcharts. Combining these flowcharts with primitive subroutines which hide most of the details of control implementation has provided an effective medium for algorithm specification and validation. Examples of the flowcharts and the language used to specify them will be given

  5. Limonene hydroperoxide analogues show specific patch test reactions.

    Science.gov (United States)

    Christensson, Johanna Bråred; Hellsén, Staffan; Börje, Anna; Karlberg, Ann-Therese

    2014-05-01

    The fragrance terpene R-limonene is a very weak sensitizer, but forms allergenic oxidation products upon contact with air. The primary oxidation products of oxidized limonene, the hydroperoxides, have an important impact on the sensitizing potency of the oxidation mixture. One analogue, limonene-1-hydroperoxide, was experimentally shown to be a significantly more potent sensitizer than limonene-2-hydroperoxide in the local lymph node assay with non-pooled lymph nodes. To investigate the pattern of reactivity among consecutive dermatitis patients to two structurally closely related limonene hydroperoxides, limonene-1-hydroperoxide and limonene-2-hydroperoxide. Limonene-1-hydroperoxide, limonene-2-hydroperoxide, at 0.5% in petrolatum, and oxidized limonene 3.0% pet. were tested in 763 consecutive dermatitis patients. Of the tested materials, limonene-1-hydroperoxide gave most reactions, with 2.4% of the patients showing positive patch test reactions. Limonene-2-hydroperoxide and oxidized R-limonene gave 1.7% and 1.2% positive patch test reactions, respectively. Concomitant positive patch test reactions to other fragrance markers in the baseline series were frequently noted. The results are in accordance with the experimental studies, as limonene-1-hydroperoxide gave more positive patch test reactions in the tested patients than limonene-2-hydroperoxide. Furthermore, the results support the specificity of the allergenic activity of the limonene hydroperoxide analogues and the importance of oxidized limonene as a cause of contact allergy. © 2014 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  6. 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)

  7. 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

  8. Efficient Noninferiority Testing Procedures for Simultaneously Assessing Sensitivity and Specificity of Two Diagnostic Tests

    Directory of Open Access Journals (Sweden)

    Guogen Shan

    2015-01-01

    Full Text Available Sensitivity and specificity are often used to assess the performance of a diagnostic test with binary outcomes. Wald-type test statistics have been proposed for testing sensitivity and specificity individually. In the presence of a gold standard, simultaneous comparison between two diagnostic tests for noninferiority of sensitivity and specificity based on an asymptotic approach has been studied by Chen et al. (2003. However, the asymptotic approach may suffer from unsatisfactory type I error control as observed from many studies, especially in small to medium sample settings. In this paper, we compare three unconditional approaches for simultaneously testing sensitivity and specificity. They are approaches based on estimation, maximization, and a combination of estimation and maximization. Although the estimation approach does not guarantee type I error, it has satisfactory performance with regard to type I error control. The other two unconditional approaches are exact. The approach based on estimation and maximization is generally more powerful than the approach based on maximization.

  9. 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

  10. 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

  11. 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

  12. 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

  13. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

  14. 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…

  15. Sport specific fitness testing of elite badminton players.

    Science.gov (United States)

    Chin, M K; Wong, A S; So, R C; Siu, O T; Steininger, K; Lo, D T

    1995-01-01

    There is a scarcity of descriptive data on the performance capacity of elite badminton players, whose fitness requirements are quite specific. The purpose of this paper is to investigate the physiological response of elite badminton players in a sport-specific fitness test. Twelve Hong Kong national badminton team players performed a field test on a badminton court. Six light bulbs were connected to a programming device causing individual bulbs to light up in a given sequence. The players were instructed to react to the flashes by running towards them, and striking shuttles mounted in the vicinity of the bulbs. Exercise intensity was controlled by altering the interval between successive lightings. A low correlation (r = 0.65) was found between the results of the field test and the rank-order list of subjects, based on an objective on-field physiological assessment and subjective ranking. This may be explained by the requirements of other factors besides physical fitness which contribute to success in elite level badminton competition. These factors may include, for example, technical skill, mental power, and aesthetic judgements on the court. Maximum mean (s.d.) heart rate data (187(8) beats.min-1) and blood lactate values (10.4(2.9) mmol.l-1) in this study showed that players were under maximal load during the field test. From the testing data, it seems reasonable to speculate that the intensity of level 3 (20 light pulses.min-1; 3.0 s.pulse-1) and level 4 (22 light pulses.min-1; 2.7 s.pulse-1) simulates the requirement of actual games energy expenditure of the Hong Kong badminton players exercising at close to their anaerobic threshold. The results also show that an estimate of fitness can be derived from measurements involving exercise closely resembling that which is specific for the sports activity in question. Improved training advice and guidance may result from such studies. PMID:8800846

  16. 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

  17. Phased Startup Initiative Phases 3 and 4 Test Plan and Test Specification (OCRWM)

    International Nuclear Information System (INIS)

    PITNER, A.L.

    2000-01-01

    Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. These tests are described in separate planning documents. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: (1) Define the test scope for the FRS and IWTS; (2) Provide detailed test requirements that can be used to write the specific test procedures; (3) Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and (4) Define specific test objectives and acceptance criteria

  18. 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.

  19. 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.

  20. 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.

  1. 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)

  2. 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)

  3. ATE accomplishes receiver specification testing with increased speed and throughput

    Science.gov (United States)

    Moser, S. A.

    1982-12-01

    The use of automatic test equipment (ATE) for receiver specifications testing can result in a 90-95% reduction of test time, with a corresponding reduction of labor costs due both to the reduction of personnel numbers and a simplification of tasks that permits less skilled personnel to be employed. These benefits free high-level technicians for more challenging system management assignments. Accuracy and repeatability also improve with the adoption of ATE, since no possibility of human error can be introduced into the readings that are taken by the system. A massive and expensive software design and development effort is identified as the most difficult aspect of ATE implementation, since programming is both time-consuming and labor intensive. An attempt is therefore made by system manufacturers to conduct an integrated development program for both ATE system hardware and software.

  4. Phase Startup Initiative Phases 3 and 4 Test Plan and Test Specification (OCRWM)

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.; LANGEVIN, M.J.

    2000-01-01

    Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: Define the Phase 3 and 4 test scope for the FRS and IWTS; Provide detailed test requirements that can be used to write the specific test procedures; Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and Define specific test objectives and acceptance criteria

  5. Phase Startup Initiative Phases 3 and 4 Test Plan and Test Specification ( OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.; LANGEVIN, M.J.

    2000-08-07

    Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: Define the Phase 3 and 4 test scope for the FRS and IWTS; Provide detailed test requirements that can be used to write the specific test procedures; Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and Define specific test objectives and acceptance criteria.

  6. Non coding RNA: sequence-specific guide for chromatin modification and DNA damage signaling

    Directory of Open Access Journals (Sweden)

    Sofia eFrancia

    2015-11-01

    Full Text Available Chromatin conformation shapes the environment in which our genome is transcribed into RNA. Transcription is a source of DNA damage, thus it often occurs concomitantly to DNA damage signaling. Growing amounts of evidence suggest that different types of RNAs can, independently from their protein-coding properties, directly affect chromatin conformation, transcription and splicing, as well as promote the activation of the DNA damage response (DDR and DNA repair. Therefore, transcription paradoxically functions to both threaten and safeguard genome integrity. On the other hand, DNA damage signaling is known to modulate chromatin to suppress transcription of the surrounding genetic unit. It is thus intriguing to understand how transcription can modulate DDR signaling while, in turn, DDR signaling represses transcription of chromatin around the DNA lesion. An unexpected player in this field is the RNA interference (RNAi machinery, which play roles in transcription, splicing and chromatin modulation in several organisms. Non-coding RNAs (ncRNAs and several protein factors involved in the RNAi pathway are well known master regulators of chromatin while only recent reports suggest that ncRNAs are involved in DDR signaling and homology-mediated DNA repair. Here, we discuss the experimental evidence supporting the idea that ncRNAs act at the genomic loci from which they are transcribed to modulate chromatin, DDR signaling and DNA repair.

  7. HYDROCOIN [HYDROlogic COde INtercomparison] Level 1: Benchmarking and verification test results with CFEST [Coupled Fluid, Energy, and Solute Transport] code: Draft report

    International Nuclear Information System (INIS)

    Yabusaki, S.; Cole, C.; Monti, A.M.; Gupta, S.K.

    1987-04-01

    Part of the safety analysis is evaluating groundwater flow through the repository and the host rock to the accessible environment by developing mathematical or analytical models and numerical computer codes describing the flow mechanisms. This need led to the establishment of an international project called HYDROCOIN (HYDROlogic COde INtercomparison) organized by the Swedish Nuclear Power Inspectorate, a forum for discussing techniques and strategies in subsurface hydrologic modeling. The major objective of the present effort, HYDROCOIN Level 1, is determining the numerical accuracy of the computer codes. The definition of each case includes the input parameters, the governing equations, the output specifications, and the format. The Coupled Fluid, Energy, and Solute Transport (CFEST) code was applied to solve cases 1, 2, 4, 5, and 7; the Finite Element Three-Dimensional Groundwater (FE3DGW) Flow Model was used to solve case 6. Case 3 has been ignored because unsaturated flow is not pertinent to SRP. This report presents the Level 1 results furnished by the project teams. The numerical accuracy of the codes is determined by (1) comparing the computational results with analytical solutions for cases that have analytical solutions (namely cases 1 and 4), and (2) intercomparing results from codes for cases which do not have analytical solutions (cases 2, 5, 6, and 7). Cases 1, 2, 6, and 7 relate to flow analyses, whereas cases 4 and 5 require nonlinear solutions. 7 refs., 71 figs., 9 tabs

  8. A test of the domain-specific acculturation strategy hypothesis.

    Science.gov (United States)

    Miller, Matthew J; Yang, Minji; Lim, Robert H; Hui, Kayi; Choi, Na-Yeun; Fan, Xiaoyan; Lin, Li-Ling; Grome, Rebekah E; Farrell, Jerome A; Blackmon, Sha'kema

    2013-01-01

    Acculturation literature has evolved over the past several decades and has highlighted the dynamic ways in which individuals negotiate experiences in multiple cultural contexts. The present study extends this literature by testing M. J. Miller and R. H. Lim's (2010) domain-specific acculturation strategy hypothesis-that individuals might use different acculturation strategies (i.e., assimilated, bicultural, separated, and marginalized strategies; J. W. Berry, 2003) across behavioral and values domains-in 3 independent cluster analyses with Asian American participants. Present findings supported the domain-specific acculturation strategy hypothesis as 67% to 72% of participants from 3 independent samples using different strategies across behavioral and values domains. Consistent with theory, a number of acculturation strategy cluster group differences emerged across generational status, acculturative stress, mental health symptoms, and attitudes toward seeking professional psychological help. Study limitations and future directions for research are discussed.

  9. 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

  10. 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)

  11. 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

  12. 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.

  13. 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.

  14. 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)

  15. Comparison of sport-specific and non-specific exercise testing in inline speed skating.

    Science.gov (United States)

    Stangier, Carolin; Abel, Thomas; Mierau, Julia; Gutmann, Boris; Hollmann, Wildor; Struder, Heiko K

    2016-04-01

    The most effective way to measure exercise performance in inline speed skating (ISS) has yet to be established. Generally most athletes are examined by means of traditional but unspecific cycling (CYC) or running (RUN) testing. The present study investigates whether a sport-specific incremental test in ISS reveals different results. Eight male top level inline speed skaters (age: 30±4 years; 65.4±6.3 mL∙kg-1∙min-1, training: 12-14 h/week) performed three incremental exhaustive tests in a randomized order (ergometer CYC, field RUN, field ISS). During the tests, heart rate (HR), oxygen uptake (V̇O2, energy expenditure (EE) and blood lactate concentration (BLC) were measured. Analysis of variance revealed no significant differences for peak HR (187±9, 191±9, 190±9; P=0.75), BLC (10.9±2.3, 10.8±2.4, 8.5±3.2; P=0.25), V̇O2 (65.4±6.3, 66.8±3.5, 66.4±6.5; P=0.91) and EE (1371±165, 1335±93, 1439±196; P=0.51) between ISS and CYC or RUN test. Similar results appeared for HR and V̇O2 at submaximal intensities (2 and 4 mmol·L-1 BLC; P≥0.05). Small to moderate effect sizes 0.3-0.87 and considerable variability of differences between the exercise modes (mean bias range between 1% and 17% with 95% limits of agreement between 3% and 33%) among submaximal and maximal results limit the comparability of the three tests. Consequently, CYC and RUN tests may be considered as qualified alternatives for a challenging ISS test. However a sport-specific test should be conducted in cases of doubt, or when precision is required (e.g. for elite athletes or scientific studies).

  16. 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

  17. Cyber-Physical Energy Systems Modeling, Test Specification, and Co-Simulation Based Testing

    DEFF Research Database (Denmark)

    van der Meer, A. A.; Palensky, P.; Heussen, Kai

    2017-01-01

    The gradual deployment of intelligent and coordinated devices in the electrical power system needs careful investigation of the interactions between the various domains involved. Especially due to the coupling between ICT and power systems a holistic approach for testing and validating is required....... Taking existing (quasi-) standardised smart grid system and test specification methods as a starting point, we are developing a holistic testing and validation approach that allows a very flexible way of assessing the system level aspects by various types of experiments (including virtual, real......, and mixed lab settings). This paper describes the formal holistic test case specification method and applies it to a particular co-simulation experimental setup. The various building blocks of such a simulation (i.e., FMI, mosaik, domain-specific simulation federates) are covered in more detail...

  18. Identity-specific coding of future rewards in the human orbitofrontal cortex.

    Science.gov (United States)

    Howard, James D; Gottfried, Jay A; Tobler, Philippe N; Kahnt, Thorsten

    2015-04-21

    Nervous systems must encode information about the identity of expected outcomes to make adaptive decisions. However, the neural mechanisms underlying identity-specific value signaling remain poorly understood. By manipulating the value and identity of appetizing food odors in a pattern-based imaging paradigm of human classical conditioning, we were able to identify dissociable predictive representations of identity-specific reward in orbitofrontal cortex (OFC) and identity-general reward in ventromedial prefrontal cortex (vmPFC). Reward-related functional coupling between OFC and olfactory (piriform) cortex and between vmPFC and amygdala revealed parallel pathways that support identity-specific and -general predictive signaling. The demonstration of identity-specific value representations in OFC highlights a role for this region in model-based behavior and reveals mechanisms by which appetitive behavior can go awry.

  19. 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.)

  20. Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Yang, T.; Cheng, X.

    2014-01-01

    To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

  1. 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

  2. Specification-based testing: What is it? How can it be automated?

    International Nuclear Information System (INIS)

    Poston, R.M.

    1994-01-01

    Software testing should begin with a written requirements specification. A specification states how software is expected to behave and describes operational characteristics (performance, reliability, etc.) for the software. A specification serves as a reference or base to test against, giving rise to the name, specification-based testing. Should analysts or designers fail to write a specification, then testers are obliged to write their own specification to test against. Specifications written by testers may be called test plans or test objectives

  3. Simulation of containment phenomena during the Phebus FPT1 test with the CONTAIN code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2002-01-01

    Thermal-hydraulic and aerosol phenomena which occurred in the containment vessel of the Phebus integral experimental facility during the first 30000 s of the Phebus FPT1 test were simulated with the CONTAIN thermal-hydraulic computer code. A single-cell input model of the vessel was developed, and boundary and initial conditions that were determined during the experiment were applied. The comparison of experimental and calculated results shows that, although the atmosphere temperature was well simulated, the calculated condensation rate was apparently too high, resulting in a lower pressure of the containment atmosphere. The aerosol deposition process was well simulated.(author)

  4. A benchmark test of computer codes for calculating average resonance parameters

    International Nuclear Information System (INIS)

    Ribon, P.; Thompson, A.

    1983-01-01

    A set of resonance parameters has been generated from known, but secret, average values; the parameters have then been adjusted to mimic experimental data by including the effects of Doppler broadening, resolution broadening and statistical fluctuations. Average parameters calculated from the dataset by various computer codes are compared with each other, and also with the true values. The benchmark test is fully described in the report NEANDC160-U (NEA Data Bank Newsletter No. 27 July 1982); the present paper is a summary of this document. (Auth.)

  5. Application of golay complementary coded excitation schemes for non-destructive testing of sandwich structures

    Science.gov (United States)

    Arora, Vanita; Mulaveesala, Ravibabu

    2017-06-01

    In recent years, InfraRed Thermography (IRT) has become a widely accepted non-destructive testing technique to evaluate the structural integrity of composite sandwich structures due to its full-field, remote, fast and in-service inspection capabilities. This paper presents a novel infrared thermographic approach named as Golay complementary coded thermal wave imaging is presented to detect disbonds in a sandwich structure having face sheets from Glass/Carbon Fibre Reinforced (GFR/CFR) laminates and core of the wooden block.

  6. Installation and testing of the ERANOS computer code for fast reactor calculations

    International Nuclear Information System (INIS)

    Gren, Milan

    2010-12-01

    The French ERANOS computer code was acquired and tested by solving benchmark problems. Five problems were calculated: 1D XZ Model, 1D RZ Model, 3D HEX SNR 300 reactor, 2S HEX and 3D HEX VVER 440 reactor. The multi-group diffuse approximation was used. The multiplication coefficients were compared within the first problem, neutron flux density in the calculation points was obtained within the second problem, and powers in the various reactor areas and in the assemblies were calculated within the remaining problems. (P.A.)

  7. ISP-50 Specifications for a Direct Vessel Injection Line Break Test with the ATLAS

    International Nuclear Information System (INIS)

    Choi, Ki Yong; Baek, Won Pil; Kim, Yeon Sik; Park, Hyun Sik; Cho, Seok; Kang, Kyoung Ho; Choi, Nam Hyun; Min, Kyoung Ho

    2009-06-01

    An OECD/NEA International Standard Problem Exercise (ISP) focussing on a DVI line break simulation result with the ATLAS was approved by the NEA Committee on the Safety of Nuclear Installation (CSNI) meeting in December 2008 and was numbered by ISP-50. The ISP-50 program will be operated by an operating agency, KAERI for three years starting from the physical year 2009. Fourteen international organizations confirmed their participation in the ISP-50, including NRC (USA), JAEA, JNES (Japan), GRS (Germany), KFKI-AEKI (Hungary), EDO Gidropress (Russia), VTT, Fortum (Finland), NRI (Czech Republic), Univ. of Pisa (Italy), KINS, KNF, KOPEC, and KAERI (Korea). In addition, KTH in Sweden and HSE in UK are considering late participation. Recently, NPIC and CIAE in China hope to join the ISP-50. As for the safety analysis codes, nine codes are expected to be used for the ISP-50: MARS-3D, RELAP5- 3D, RELAP5, TRACE, CATHARE, APROS, ATHELET, TRAP, and KORSAR. It is the first ISP exercise in Korea in which a domestic test facility is utilized by international nuclear society and this exercise will contribute to extending our physical understanding on thermal hydraulic phenomena during the DVI line break accidents and to verifying the best-estimate thermal-hydraulic safety analysis codes. This report was prepared to define technical specifications of the ISP-50 exercise according the guideline provided by OECD/CSNI. It includes general objectives, phases, deliverables to participants, parameters required for comparison and the time table

  8. Improving Inpatient Surveys: Web-Based Computer Adaptive Testing Accessed via Mobile Phone QR Codes.

    Science.gov (United States)

    Chien, Tsair-Wei; Lin, Weir-Sen

    2016-03-02

    The National Health Service (NHS) 70-item inpatient questionnaire surveys inpatients on their perceptions of their hospitalization experience. However, it imposes more burden on the patient than other similar surveys. The literature shows that computerized adaptive testing (CAT) based on item response theory can help shorten the item length of a questionnaire without compromising its precision. Our aim was to investigate whether CAT can be (1) efficient with item reduction and (2) used with quick response (QR) codes scanned by mobile phones. After downloading the 2008 inpatient survey data from the Picker Institute Europe website and analyzing the difficulties of this 70-item questionnaire, we used an author-made Excel program using the Rasch partial credit model to simulate 1000 patients' true scores followed by a standard normal distribution. The CAT was compared to two other scenarios of answering all items (AAI) and the randomized selection method (RSM), as we investigated item length (efficiency) and measurement accuracy. The author-made Web-based CAT program for gathering patient feedback was effectively accessed from mobile phones by scanning the QR code. We found that the CAT can be more efficient for patients answering questions (ie, fewer items to respond to) than either AAI or RSM without compromising its measurement accuracy. A Web-based CAT inpatient survey accessed by scanning a QR code on a mobile phone was viable for gathering inpatient satisfaction responses. With advances in technology, patients can now be offered alternatives for providing feedback about hospitalization satisfaction. This Web-based CAT is a possible option in health care settings for reducing the number of survey items, as well as offering an innovative QR code access.

  9. Towards sensible toxicity testing for nanomaterials: proposal for the specification of test design

    International Nuclear Information System (INIS)

    Potthoff, Annegret; Meißner, Tobias; Weil, Mirco; Kühnel, Dana

    2015-01-01

    During the last decade, nanomaterials (NM) were extensively tested for potential harmful effects towards humans and environmental organisms. However, a sound hazard assessment was so far hampered by uncertainties and a low comparability of test results. The reason for the low comparability is a high variation in the (1) type of NM tested with regard to raw material, size and shape and (2) procedures before and during the toxicity testing. This calls for tailored, nanomaterial-specific protocols. Here, a structured approach is proposed, intended to lead to test protocols not only tailored to specific types of nanomaterials, but also to respective test system for toxicity testing. There are existing standards on single procedures involving nanomaterials, however, not all relevant procedures are covered by standards. Hence, our approach offers a detailed way of weighting several plausible alternatives for e.g. sample preparation, in order to decide on the procedure most meaningful for a specific nanomaterial and toxicity test. A framework of several decision trees (DT) and flow charts to support testing of NM is proposed as a basis for further refinement and in-depth elaboration. DT and flow charts were drafted for (1) general procedure—physicochemical characterisation, (2) choice of test media, (3) decision on test scenario and application of NM to liquid media, (4) application of NM to the gas phase, (5) application of NM to soil and sediments, (6) dose metrics, (S1) definition of a nanomaterial, and (S2) dissolution. The applicability of the proposed approach was surveyed by using experimental data retrieved from studies on nanoscale CuO. This survey demonstrated the DT and flow charts to be a convenient tool to systematically decide upon test procedures and processes, and hence pose an important step towards harmonisation of NM testing. (paper)

  10. 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.

  11. Power-Cooling-Mismatch Test Series Test PCM-7. Experiment operating specifications

    International Nuclear Information System (INIS)

    Sparks, D.T.; Smith, R.H.; Stanley, C.J.

    1979-02-01

    The experiment operating specifications for the Power-Cooling-Mismatch (PCM) Test PCM-7 to be conducted in the Power Burst Facility are described. The PCM Test Series was designed on the basis of a parametric evaluation of fuel behavior response with cladding temperature, rod internal pressure, time in film boiling, and test rod power being the variable parameters. The test matrix, defined in the PCM Experiment Requirements Document (ERD), encompasses a wide range of situations extending from pre-CHF (critical heat flux) PCMs to long duration operation in stable film boiling leading to rod failure

  12. Pump Component Model in SPACE Code

    International Nuclear Information System (INIS)

    Kim, Byoung Jae; Kim, Kyoung Doo

    2010-08-01

    This technical report describes the pump component model in SPACE code. A literature survey was made on pump models in existing system codes. The models embedded in SPACE code were examined to check the confliction with intellectual proprietary rights. Design specifications, computer coding implementation, and test results are included in this report

  13. Outlines and verifications of the codes used in the safety analysis of High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Kunitomi, Kazuhiko; Maruyama, Soh; Fujita, Shigeki; Nakagawa, Shigeaki; Iyoku, Tatsuo; Shindoh, Masami; Sudo, Yukio; Hirano, Masashi.

    1990-03-01

    This paper presents brief description of the computer codes used in the safety analysis of High Temperature Engineering Test Reactor. The list of the codes is: 1. BLOOST-J2 2. THYDE-HTGR 3. TAC-NC 4. RATSAM6 5. COMPARE-MOD1 6. GRACE 7. OXIDE-3F 8. FLOWNET/TRUMP. Of described above, 1, 3, 4, 5, 6 and 7 were developed for the multi-hole type gas cooled reactor and improved for HTTR and 2 was originated by THYDE-codes which were developed to treat the transient thermo-hydraulics during LOCA of LWR. Each code adopted the models and properties which yield conservative analytical results. Adequacy of each code was verified by the comparison with the experimental results and/or the analytical results obtained from the other codes which were already proven. (author)

  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. Towards a European draft code of practice in creep crack growth testing

    International Nuclear Information System (INIS)

    Nikbin, K.M.

    2003-01-01

    Crack growth and initiation models as well as defect assessment codes need reliable and verifiable material properties data for use in their predictive methodologies. These data consist of uniaxial, multiaxial and crack initiation and growth data under static and cyclic loading at the relevant temperatures. International collaboration for developing standards in this field started in 1987 under the auspices of the VAMAS (Versailles Agreement for MAterials and Standards). Two technical Working Areas TWA11 and TWA19 committees ending 1998 have made substantial progress in unifying and standardising the methods for obtaining the relevant data. This collaboration has resulted in the development of ASTM E1457 creep crack growth testing standard. The European collaborative programme CRETE (see Acknowledgements), which began in 1999, is following up this valuable research in order to develop a European Code of Practice for elevated temperature crack growth which is planned to have a wider field of application. A Round Robin experimental, analytical and verification programme in CRETE will include testing a type 316 LN stainless steel at 550 degC and a Carbon-Manganese steel at 400 degC consisting of seven different geometries. The paper reviews the methods of analysis used for laboratory creep crack growth data and their relevance to long term crack initiation and growth in components. In addition, since design and life assessment and material properties under creep are an integral part of this project a short review of the models available for predicting creep and fatigue crack growth is presented. (author)

  16. Specific long non-coding RNAs response to occupational PAHs exposure in coke oven workers

    Directory of Open Access Journals (Sweden)

    Chen Gao

    Full Text Available To explore whether the alteration of lncRNA expression is correlated with polycyclic aromatic hydrocarbons (PAHs exposure and DNA damage, we examined PAHs external and internal exposure, DNA damage and lncRNAs (HOTAIR, MALAT1, TUG1 and GAS5 expression in peripheral blood lymphocytes (PBLCs of 150 male coke oven workers and 60 non-PAHs exposure workers. We found the expression of HOTAIR, MALAT1, and TUG1 were enhanced in PBLCs of coke oven workers and positively correlated with the levels of external PAHs exposure (adjusted Ptrend < 0.001 for HOTAIR and MALAT1, adjusted Ptrend = 0.006 for TUG1. However, only HOTAIR and MALAT1 were significantly associated with the level of internal PAHs exposure (urinary 1-hydroxypyrene with adjusted β = 0.298, P = 0.024 for HOTAIR and β = 0.090, P = 0.034 for MALAT1. In addition, the degree of DNA damage was positively associated with MALAT1 and HOTAIR expression in PBLCs of all subjects (adjusted β = 0.024, P = 0.002 for HOTAIR and β = 0.007, P = 0.003 for MALAT1. Moreover, we revealed that the global histone 3 lysine 27 trimethylation (H3K27me3 modification was positively associated with the degree of genetic damage (β = 0.061, P < 0.001 and the increase of HOTAIR expression (β = 0.385, P = 0.018. Taken together, our findings suggest that altered HOTAIR and MALAT1 expression might be involved in response to PAHs-induced DNA damage. Keywords: Polycyclic aromatic hydrocarbons, Long non-coding RNA, Peripheral blood lymphocytes, DNA damage response, HOTAIR, MALAT

  17. Tests of a 3D Self Magnetic Field Solver in the Finite Element Gun Code MICHELLE

    CERN Document Server

    Nelson, Eric M

    2005-01-01

    We have recently implemented a prototype 3d self magnetic field solver in the finite-element gun code MICHELLE. The new solver computes the magnetic vector potential on unstructured grids. The solver employs edge basis functions in the curl-curl formulation of the finite-element method. A novel current accumulation algorithm takes advantage of the unstructured grid particle tracker to produce a compatible source vector, for which the singular matrix equation is easily solved by the conjugate gradient method. We will present some test cases demonstrating the capabilities of the prototype 3d self magnetic field solver. One test case is self magnetic field in a square drift tube. Another is a relativistic axisymmetric beam freely expanding in a round pipe.

  18. Development of European creep crack growth testing code of practice for industrial specimens

    Energy Technology Data Exchange (ETDEWEB)

    Dogan, B.; Nikbin, K. [Imperial College, London (United Kingdom); Petrovski, B. [Technische Univ. Darmstadt (DE). Inst. fuer Werkstoffkunde (IFW)

    2004-07-01

    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 elasticcreep, elastic compliance measurements.

  19. Deliverable 3.3.2 Specification of tests and test groups

    DEFF Research Database (Denmark)

    Peterson, Carrie Beth; Mitseva, Anelia; Harpur, Jill

    2009-01-01

    Deliverable 3.3.2: Specification of tests and test groups One of the main goals of the ISISEMD project is to offer innovative ICT services to improve the quality of life of elderly persons with cognitive problems or mild dementia and their informal and formal caregivers who provide every day care...... for them. This will be done via integrating intelligent scalable ICT services which will be tested for a period of 12 months under realistic conditions. Offering the services could not be complete without evaluating quality of life improvement, user acceptance and user satisfaction with a representative...... group of the target user groups. This document is devoted to describing important aspects of services evaluation such as: who the test participants will be, inclusion and exclusion criterion, selection standards, how the test participants will be recruited, ethical considerations, etc. Test methodology...

  20. Assessment of the MARS-KS Code Using Atlas 6-inch cold leg Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kang, D. G.; Kim, J. S.; Ahn, S. H.; Seul, K. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-03-15

    An integral effect test on the SBLOCA (Small-Break Loss of Coolant Accident) aiming at 6-inch cold leg bottom break, SB-CL-09, was conducted with the Atlas on November, 13, 2009, by KAERI. In this study, the calculations using MARS-KS Vt1.2 code were conducted for 6-inch cold leg break test of Atlas (SB-CL-09) which is the second domestic standard problem (Dsp-02) to assess MARS-KS code capability to simulate the transient thermal-hydraulic behavior for SBLOCA. The steady state was determined by conducting a null transient calculation and the errors between the calculated and measured values are acceptable for almost primary/secondary system parameters. The predicted pressurizer pressure agrees relatively well with the experimental data and the predicted break flow and mass are in good agreement with experiment. In MARS-KS calculation, the decrease of core collapsed water level is predicted well in blowdown phase, but just before LSC, water level is higher than experiment. However, the sudden decrease and increase of water level is higher than experiment. However, the sudden decrease and increase of water level at the LSC are predicted qualitatively. After LSC, there is another water level dip at Sit injection time which is not in experiment. It is considered that this phenomenon is caused by rapid depressurization of downcomer due to significant condensation rate of vapor in downcomer when Sit water flows in it. For the downcomer water level is predicted well, however, it is significantly over-predicted at SIT injection time, water level is predicted well, however, it is significantly over-predicted at SIT injection time after SIT water flows in downcomer. Predicted cladding temperature generally agrees well with the experiment, while there is peak at SIT injection time in calculation which is not in experiment. The loop seals of 1A, 2B intermediate leg are cleared around 400 seconds in experiment, while only that of 1A is cleared in MARS-KS calculation at the

  1. 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

  2. Results and code prediction comparisons of lithium-air reaction and aerosol behavior tests

    International Nuclear Information System (INIS)

    Jeppson, D.W.

    1986-03-01

    The Hanford Engineering Development Laboratory (HEDL) Fusion Safety Support Studies include evaluation of potential safety and environmental concerns associated with the use of liquid lithium as a breeder and coolant for fusion reactors. Potential mechanisms for volatilization and transport of radioactive metallic species associated with breeder materials are of particular interest. Liquid lithium pool-air reaction and aerosol behavior tests were conducted with lithium masses up to 100 kg within the 850-m 3 containment vessel in the Containment Systems Test Facility. Lithium-air reaction rates, aerosol generation rates, aerosol behavior and characterization, as well as containment atmosphere temperature and pressure responses were determined. Pool-air reaction and aerosol behavior test results were compared with computer code calculations for reaction rates, containment atmosphere response, and aerosol behavior. The volatility of potentially radioactive metallic species from a lithium pool-air reaction was measured. The response of various aerosol detectors to the aerosol generated was determined. Liquid lithium spray tests in air and in nitrogen atmospheres were conducted with lithium temperatures of about 427 0 and 650 0 C. Lithium reaction rates, containment atmosphere response, and aerosol generation and characterization were determined for these spray tests

  3. 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.

  4. Specifications for a two-dimensional multi-group scattering code: ALCI

    International Nuclear Information System (INIS)

    Bayard, J.P.; Guillou, A.; Lago, B.; Bureau du Colombier, M.J.; Guillou, G.; Vasseur, Ch.

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [fr

  5. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  6. Post test calculation of the experiment 'small break loss-of- coolant test' SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    International Nuclear Information System (INIS)

    Lischke, W.; Vandreier, B.

    1997-01-01

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory

  7. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W; Vandreier, B [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1998-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  8. Coupling Hydrodynamic and Wave Propagation Codes for Modeling of Seismic Waves recorded at the SPE Test.

    Science.gov (United States)

    Larmat, C. S.; Rougier, E.; Delorey, A.; Steedman, D. W.; Bradley, C. R.

    2016-12-01

    The goal of the Source Physics Experiment (SPE) is to bring empirical and theoretical advances to the problem of detection and identification of underground nuclear explosions. For this, the SPE program includes a strong modeling effort based on first principles calculations with the challenge to capture both the source and near-source processes and those taking place later in time as seismic waves propagate within complex 3D geologic environments. In this paper, we report on results of modeling that uses hydrodynamic simulation codes (Abaqus and CASH) coupled with a 3D full waveform propagation code, SPECFEM3D. For modeling the near source region, we employ a fully-coupled Euler-Lagrange (CEL) modeling capability with a new continuum-based visco-plastic fracture model for simulation of damage processes, called AZ_Frac. These capabilities produce high-fidelity models of various factors believed to be key in the generation of seismic waves: the explosion dynamics, a weak grout-filled borehole, the surrounding jointed rock, and damage creation and deformations happening around the source and the free surface. SPECFEM3D, based on the Spectral Element Method (SEM) is a direct numerical method for full wave modeling with mathematical accuracy. The coupling interface consists of a series of grid points of the SEM mesh situated inside of the hydrodynamic code's domain. Displacement time series at these points are computed using output data from CASH or Abaqus (by interpolation if needed) and fed into the time marching scheme of SPECFEM3D. We will present validation tests with the Sharpe's model and comparisons of waveforms modeled with Rg waves (2-8Hz) that were recorded up to 2 km for SPE. We especially show effects of the local topography, velocity structure and spallation. Our models predict smaller amplitudes of Rg waves for the first five SPE shots compared to pure elastic models such as Denny &Johnson (1991).

  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. Hydraulic Simulation of In-vessel Downstream Effect Test Using MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Lee, Joon Soo; Ryu, Seung Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    In-vessel downstream effect test (IDET) has been required to evaluate the effect of debris on long term core cooling following a loss of coolant accident (LOCA) in support of resolution of Generic Safety Issue (GSI) 191. Head loss induced by debris (fiber and particle) accumulated on prototypical fuel assembly (FA) should be compared with the available driving head to the core for the various combinations of LOCA and Emergency Core Cooling System (ECCS) injection. The actual simulation was conducted using MARS-KS code. Also the influence of small difference in gap size which was found in the actual experiment is evaluated using the present model. A simple model to determine the form loss factors of FA and gap in clean state and the debris laden state is discussed based on basic fluid mechanics. Those form loss factors were applied to the hydraulic simulation of a selected IDET using MARS-KS code. The result indicated that the present model can be applied to IDET simulation. The pressure drop influenced by small difference in gap size can be evaluated by the present model with practical assumption.

  12. 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)

  13. 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)

  14. Verification and validation of the PLTEMP/ANL code for thermal hydraulic analysis of experimental and test reactors

    International Nuclear Information System (INIS)

    Kalimullah, M.; Olson, A.O.; Feldman, E.E.; Hanan, N.; Dionne, B.

    2012-01-01

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  15. Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-07

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  16. Specific binding-adsorbent assay method and test means

    International Nuclear Information System (INIS)

    1981-01-01

    A description is given of an improved specific binding assay method and test means employing a nonspecific adsorbent for the substance to be determined, particularly hepatitis B surface (HBsub(s)) antigen, in its free state or additionally in the form of its immune complex. The invention is illustrated by 1) the radioimmunoadsorbent assay for HBsub(s) antigen, 2) the radioimmunoadsorbent assay for HBsub(s) antigen in the form of immune complex with antibody, 3) a study of adsorption characteristics of various anion exchange materials for HBsub(s) antigen, 4) the use of hydrophobic adsorbents in a radioimmunoadsorbent assay for HBsub(s) antigen and 5) the radioimmunoadsorbent assay for antibody to HBsub(s) antigen. The advantages of the present method for detecting HBsub(s) antigen compared to previous methods include the manufacturing advantages of eliminating the need for insolubilised anti-HBsub(s) and the advantages of a single incubation step, fewer manipulations, storability of adsorbent materials, increased sensitivity and versatility of detecting HBsub(s) antigen in the form of its immune complex if desired. (U.K.)

  17. Specification and acceptance testing of nuclear medicine equipment

    International Nuclear Information System (INIS)

    Wegst, A.V.; Erickson, J.J.

    1984-01-01

    The purchase of nuclear medicine equipment is of prime importance in the operation of a clinical service. Failure to properly evaluate the potential uses of the instrumentation and the various operational characteristics of the equipment can often result in the purchase of inappropriate or inferior instruments. The magnitude of the purchase in terms of time and financial investments make it imperative that the purchase be approached in a systematic manner. Consideration of both the intended clinical functions and personnel requirements is important. It is necessary also to evaluate the ability of the equipment vendor to support the instrumentation after the purchase has been completed and the equipment installed in the clinical site. The desired specifications of the instrument characteristics should be stated in terms that can be verified by acceptance testing. The complexity of modern instrumentation and the sensitivity of it to the environment require the buyer to take into account the potential problems of controlling the temperature, humidity, and electrical power of the installation site. If properly and systematically approached, the purchase of new nuclear medicine instrumentation can result in the acquisition of a powerful diagnostic tool which will have a useful lifetime of many years. If not so approached, it may result in the expenditure of a large amount of money and personnel time without the concomitant return in useful clinical service. (author)

  18. MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-06-12

    This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.

  19. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  20. Preliminary code development for seismic signal analysis related to test ban treaty questions

    International Nuclear Information System (INIS)

    Brolley, J.E.

    1977-01-01

    Forensic seismology, from a present day viewpoint, appears to be divided into several areas. Overwhelmingly important, in view of current Complete Test Ban (CTB) discussions, is the seismological study of waves generated in the earth by underground nuclear explosions. Over the last two decades intensive effort has been devoted to developing improved observational apparatus and to the interpretation of the data produced by this equipment. It is clearly desirable to extract the maximum amount of information from seismic signals. It is, therefore, necessary to quantitatively compare various modes of analysis to establish which mode or combination of modes provides the most useful information. Preliminary code development for application of some modern developments in signal processing to seismic signals is described. Applications of noncircular functions are considered and compared with circular function results. The second portion of the discussion concerns maximum entropy analysis. Lastly, the multivariate aspects of the general problem are considered

  1. Calculation of neutron spectra produced in neutron generator target: Code testing.

    Science.gov (United States)

    Gaganov, V V

    2018-03-01

    DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. Assessment of SPACE code for multiple failure accident: 1% Cold Leg Break LOCA with HPSI failure at ATLAS Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Hyuk; Lee, Seung Wook; Kim, Kyung-Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Design extension conditions (DECs) is a popular key issue after the Fukushima accident. In a viewpoint of the reinforcement of the defense in depth concept, a high-risk multiple failure accident should be reconsidered. The target scenario of ATLAS A5.1 test was LSTF (Large Scale Test Facility) SB-CL-32 test, a 1% SBLOCA with total failure of high pressure safety injection (HPSI) system of emergency core cooling system (ECCS) and secondary side depressurization as the accident management (AM) action, as a counterpart test. As the needs to prepare the DEC accident because of a multiple failure of the present NPPs are emphasized, the capability of SPACE code, just like other system analysis code, is required to expand the DEC area. The objectives of this study is to validate the capability of SPACE code for a DEC scenario, which represents multiple failure accident like as a SBLOCA with HPSI fail. Therefore, the ATLAS A5.1 test scenario was chosen. As the needs to prepare the DEC accident because of a multiple failure of operating NPPs are emphasized, the capability of SPACE code is needed to expand the DEC area. So the capability of SPACE code was validated for one of a DEC scenario. The target scenario was selected as the ATLAS A5.1 test, which is a 1% SBLOCA with total failure of HPSI system of ECCS and secondary side depressurization. Through the sensitivity study on discharge coefficient of break flow, the best fit of integrated mass was found. Using the coefficient, the ATLAS A5.1 test was analyzed using the SPACE code. The major thermal hydraulic parameters such as the system pressure, temperatures were compared with the test and have a good agreement. Through the simulation, it was concluded that the SPACE code can effectively simulate one of multiple failure accidents like as SBLOCA with HPSI failure accident.

  3. Application of the source term code package to obtain a specific source term for the Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Souto, F.J.

    1991-06-01

    The main objective of the project was to use the Source Term Code Package (STCP) to obtain a specific source term for those accident sequences deemed dominant as a result of probabilistic safety analyses (PSA) for the Laguna Verde Nuclear Power Plant (CNLV). The following programme has been carried out to meet this objective: (a) implementation of the STCP, (b) acquisition of specific data for CNLV to execute the STCP, and (c) calculations of specific source terms for accident sequences at CNLV. The STCP has been implemented and validated on CDC 170/815 and CDC 180/860 main frames as well as on a Micro VAX 3800 system. In order to get a plant-specific source term, data on the CNLV including initial core inventory, burn-up, primary containment structures, and materials used for the calculations have been obtained. Because STCP does not explicitly model containment failure, dry well failure in the form of a catastrophic rupture has been assumed. One of the most significant sequences from the point of view of possible off-site risk is the loss of off-site power with failure of the diesel generators and simultaneous loss of high pressure core spray and reactor core isolation cooling systems. The probability for that event is approximately 4.5 x 10 -6 . This sequence has been analysed in detail and the release fractions of radioisotope groups are given in the full report. 18 refs, 4 figs, 3 tabs

  4. 7 CFR 51.3417 - Optional test for specific gravity.

    Science.gov (United States)

    2010-01-01

    ... be corrected for temperature variations using Table I. (2) A hydrometer specifically designed for determining the specific gravity of potatoes. 3 3 The hydrometer is available from the Potato Chip/Snack Food...

  5. 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.

  6. An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

    1991-05-01

    The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

  7. 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.

  8. CSNI Integral test facility validation matrix for the assessment of thermal-hydraulic codes for LWR LOCA and transients

    International Nuclear Information System (INIS)

    1996-07-01

    This report deals with an internationally agreed integral test facility (ITF) matrix for the validation of best estimate thermal-hydraulic computer codes. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a life of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of such a matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated around the world over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case

  9. 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.)

  10. Benchmark test of MORSE-DD code using double-differential form cross sections

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa; Ishiguro, Yukio

    1985-02-01

    The multi-group double-differential form cross sections (DDX) and the three dimensional Monte Carlo code MORSE-DD devised to utilize the DDX, which were developed for the fusion neutronics analysis, have been validated through many benchmark tests. All the problems tested have a 14 MeV neutron source. To compare the calculated results with the measured values, the following experiments were adopted as the benchmark problems; leakage neutron spectra from spheres composed of nine kinds of materials measured at LLNL, neutron angular spectra from the Li 2 O slab measured at FNS in JAERI, tritium production rate (TPR) in the graphite-reflected Li 2 O sphere measured at FNS and the TPR in the metallic Li sphere measured at KfK. In addition in order to test an accuracy of the calculation method in detail, spectra of neutrons scattered from a small sample and various reaction rates in a Li 2 O cylinder were compared between the present method and the continuous energy Monte Carlo method. The nuclear data files used are mainly ENDF/B4 and partly JENDL-3PR1. The tests were carried out through a comparison with the measured values and also with the results obtained from the conventional Legendre expansion method and the continuous energy Monte Carlo method. It is found that the results by the present method are more accurate than those by the conventional one and agree well with those by the continuous energy Monte Carlo calculations. Discrepancies due to the nuclear data are also discussed. (author)

  11. 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

  12. 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)

  13. Technical specification upgrading at the Fast Flux Test Facility (FFTF)

    International Nuclear Information System (INIS)

    Baird, Q.L.; Franz, G.R.; Absher, K.R.

    1985-01-01

    The FFTF Technical Specifications were generated in 1977 and 1978 following submittal of the FSAR in 1976. A phased implementation program served to prepare the specifications for each stage of the plant startup with the complete specifications approved and implemented late in 1980 for the first ascent to full power. In January, 1983 WHC undertook an upgrading effort to implement changes to the FFTF technical specifications. This program has been pursued with appropriate attention to the CFR and industry standards and practice. Examples of these changes, discussion of the methods and planned activities for the future will be presented. Technical data will be provided to support the impact of specific limits. The benefits of changes and the criteria for change will be elaborated

  14. 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

  15. Specifications, quality control, manufacturing, and testing of accelerator magnets

    CERN Document Server

    Einfeld, D

    2010-01-01

    The performance of the magnets plays an important role in the functioning of an accelerator. Most of the magnets are designed at the accelerator laboratory and built by industry. The link between the laboratory and the manufacturer is the contract containing the Technical Specifications of the magnets. For an overview of the contents of the Technical Specifications, the specifications for the magnets of ALBA (bending, quadrupole, and sextupole) are described in this paper. The basic rules of magnet design are reviewed in Appendix A.

  16. Sensitivity and specificity of the nickel spot (dimethylglyoxime) test

    DEFF Research Database (Denmark)

    Thyssen, Jacob P; Skare, Lizbet; Lundgren, Lennart

    2010-01-01

    The accuracy of the dimethylglyoxime (DMG) nickel spot test has been questioned because of false negative and positive test reactions. The EN 1811, a European standard reference method developed by the European Committee for Standardization (CEN), is fine-tuned to estimate nickel release around...... the limit value of the EU Nickel Directive from products intended to come into direct and prolonged skin contact. Because assessments according to EN 1811 are expensive to perform, time consuming, and may destruct the test item, it should be of great value to know the accuracy of the DMG screening test....

  17. 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.

  18. 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

  19. Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

    International Nuclear Information System (INIS)

    Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

    2017-01-01

    Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

  20. The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Karbojian, Aram, E-mail: karbojan@kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2015-08-15

    Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  1. Sensitivity and specificity of the nickel spot (dimethylglyoxime) test

    DEFF Research Database (Denmark)

    Thyssen, Jacob P; Skare, Lizbet; Lundgren, Lennart

    2010-01-01

    The accuracy of the dimethylglyoxime (DMG) nickel spot test has been questioned because of false negative and positive test reactions. The EN 1811, a European standard reference method developed by the European Committee for Standardization (CEN), is fine-tuned to estimate nickel release around...

  2. Codes in the codons: construction of a codon/amino acid periodic table and a study of the nature of specific nucleic acid-protein interactions.

    Science.gov (United States)

    Benyo, B; Biro, J C; Benyo, Z

    2004-01-01

    The theory of "codon-amino acid coevolution" was first proposed by Woese in 1967. It suggests that there is a stereochemical matching - that is, affinity - between amino acids and certain of the base triplet sequences that code for those amino acids. We have constructed a common periodic table of codons and amino acids, where the nucleic acid table showed perfect axial symmetry for codons and the corresponding amino acid table also displayed periodicity regarding the biochemical properties (charge and hydrophobicity) of the 20 amino acids and the position of the stop signals. The table indicates that the middle (2/sup nd/) amino acid in the codon has a prominent role in determining some of the structural features of the amino acids. The possibility that physical contact between codons and amino acids might exist was tested on restriction enzymes. Many recognition site-like sequences were found in the coding sequences of these enzymes and as many as 73 examples of codon-amino acid co-location were observed in the 7 known 3D structures (December 2003) of endonuclease-nucleic acid complexes. These results indicate that the smallest possible units of specific nucleic acid-protein interaction are indeed the stereochemically compatible codons and amino acids.

  3. 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.

  4. International standard problem (ISP) No. 43 Rapid boron-dilution transient tests for code verification. Comparison report

    International Nuclear Information System (INIS)

    2001-03-01

    International Standard Problem No. 43 (ISP 43) addresses the nuclear industries present capabilities of simulating fluid dynamics aspects of a subset of rapid boron dilution transients. Specifically, the exercise focuses on the sequence involving the transport of a boron-dilute slug through the actuation of a pump. The slug is formed on the primary side of the steam generator as a consequence of in interfacing system leak from the secondary un-borated coolant. Experimental data was collected using the University of Maryland 2 x 4 Thermalhydraulic Loop (UM 2 x 4 Loop) and the Boron-mixing Visualization Facility. Two blind test series were proposed during the first workshop (October 1998) and refined using participant input. The first series, test series A, deals with the injection of a front, i.e., a single interface between borated and dilute fluids. The second blind series, test series B, is the more realistic injection of a slug, i.e., a dilute fluid volume preceded and followed by the borated coolant of the primary system. Data are collected in the UM 2 x 4 Loop and refined details are obtained from the Visualization Facility, which represents a replica of the Loop.s vessel downcomer. In the Loop experimental program, the dilute volume is simulated by cold water and the borated primary coolant is simulated by hot water. The Visualization Facility uses dye to mark the diluted front or slug. The measured boundary conditions for both test series include the initial temperature of the primary system, the front/slug injection flowrate and temperature, and the pressure drop across the core. Temperature data is collected at 185 thermocouple positions in the downcomer and 38 positions in the lower plenum. The advancement of the front/slug through the system is monitored at discrete horizontal levels that contain the thermocouples. The performance of codes is measured relative to a set of figures of merit. During the first workshop, the principal figure of merit was

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

    NARCIS (Netherlands)

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

    2008-01-01

    This paper represents an extension to our previous work: Mining software repositories to study coevolution of production & test code. Proceedings of the International Conference on Software Testing, Verification, and Validation (ICST), IEEE Computer Society, 2008; doi:10.1109/ICST.2008.47

  6. Specifications for Testing Procedures at Svåheia

    DEFF Research Database (Denmark)

    Margheritini, Lucia; Kofoed, Jens Peter

    This report is realized in order to define testing procedures of the SSG pilot in Svåheia location. This will be done by listing all the relevant parameters related to the structure performance.......This report is realized in order to define testing procedures of the SSG pilot in Svåheia location. This will be done by listing all the relevant parameters related to the structure performance....

  7. Sensitivity and specificity of copper sulphate test in determining ...

    African Journals Online (AJOL)

    Background: The accuracy of the copper sulphate method for the rapid screening of prospective blood donors has been questioned because this rapid screening method may lead to false deferral of truly eligible prospective blood donors. Objective: This study was aimed at determining the sensitivity and specificity of copper ...

  8. A Simulation Study about OECD-SETH PANDA Tests by using MARS Code

    International Nuclear Information System (INIS)

    Bae, Sung Won; Chung, Bub Dong

    2007-04-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. In addition, the multi-D module component has been developed to meet the expand the multi-dimensional analysis capability of MARS. Participating in OECD-SETH, MARS provides and undergoes the assess procedure of comercial CFD codes, like FLUENT, CFX, etc. During the participation, MARS has been used to provide the system code results, which is made with the intermediate length scale, restricted analysis volume numbers. With these restrictions and shortcomings, MARS predicts well about the steam concentration distribution and mixture temperature in the large multi-comparted bulk spaces. After the SETH project, NEA has planned the SETH II, which deals with the multiple non-condensible gas stratification and mixing phenomena

  9. Acceptance and validation test report for HANSF code version 1.3.2

    International Nuclear Information System (INIS)

    PIEPHO, M.G.

    2001-01-01

    The HANSF code, Version 1.3.2, is a stand-along code that runs only in DOS. As a result, it runs on any Windows' platform, since each Windows(trademark) platform can create a DOS-prompt window and execute HANSF in the DOS window. The HANSF code is proprietary to Fauske and Associates, Inc., (FAI) of Burr Ridge, IL, the developers of the code. The SNF Project has a license from FAI to run the HANSF code on any computer for only work related to SNF Project. The SNF Project owns the MCO.FOR routine, which is the main routine in HANSF for CVDF applications. The HANSF code calculates physical variables such as temperature, pressure, oxidation rates due to chemical reactions of uranium metal/fuel with water or oxygen. The code is used by the Spent Nuclear Fuel (SNF) Project at Hanford; for example, the report Thermal Analysis of Cold Vacuum Drying of Spent Nuclear Fuel (HNF-SD-SNF-CN-023). The primary facilities of interest are the K-Basins, Cold Vacuum Drying Facility (CVDF), Canister Storage Building (CSB) and T Plant. The overall Summary is presented in Section 2.0, Variances in Section 3.0, Comprehensive Assessment in Section 4.0, Results in Section 5.0, Evaluation in Section 6.0, and Summary of Activities in Section 7.0

  10. Imouraren - uranium leaching tests and specificities with analcites

    International Nuclear Information System (INIS)

    Wattinne-Morice, A.; Belieres, M.

    2010-01-01

    Imouraren is a sedimentary uranium deposit (total > 150 000 tU, average U ~ 0.08 %), located in Niger (~ 100 km from Agadez). Uranium mineralization is trapped in sandstones and is widely oxidized (uranotyle, metatuyamunite), but a part remains reduced (pitchblende, uraninite). The sandstones have a peculiar mineralogical assemblage (analcite partly chloritized) which can affect uranium recovery. Several acid heap leaching tests have been completed to determine the most suitable process parameters. Microscopic studies and XRD analysis performed on fresh ore and on leached residue highlight the complex behavior of uranium and the associated mineralogical families during the tests. (author)

  11. English-Japanese terms in nondestructive testing specifications

    International Nuclear Information System (INIS)

    1991-01-01

    For technical development, it is the prerequisite to clarify the terms to be used in various fields and their definition, therefore, in various foreign countries, there are some standards on terms in respective fields, and also as international standards, ISO/TC 135 (Non-destructive testing) organized the SC on terms from the beginning of foundation. In JIS, there is the column for corresponding English (for reference), but there is the problem of English and American English. The English used in ISO, BS or EN and ASTM standards in relation to nondestructive testing were collected in every technical field and put in order, and the corresponding English terms were selected. Moreover at this opportunity, the terms having the definition in these international, national and semi-national standards were classified into eight fields, that is, common (approval, quality assurance, defects and others), radiography, ultrasonic flaw detection, acoustic emission, eddy current flaw detection, magnetic flaw detection, liquid penetrant testing and leak test, and the Japanese translation was stipulated. The draft of this standard was approved by the standardization committee on January 17, 1991. (K.I.)

  12. Evaluation of a specific test in cross-country skiing

    DEFF Research Database (Denmark)

    Mygind, Erik; Larsson, Benny; Klausen, Tom

    1991-01-01

    -poling was correlated with performance, expressed as a ranking score during 10 ski races. The tests were undertaken in September, December and April. Upper body maximal oxygen uptake increased 5.8% from September to December, decreasing to 2.3% above the September level in April. Upper body work output (2 min...

  13. 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

  14. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  15. 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

  16. Comparison of the Aerospace Systems Test Reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1983-01-01

    This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10 4 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code

  17. Comparison of the aerospace systems test reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1984-01-01

    This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10 4 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code. (author)

  18. Implicit and semi-implicit schemes in the Versatile Advection Code : numerical tests

    NARCIS (Netherlands)

    Tóth, G.; Keppens, R.; Bochev, Mikhail A.

    1998-01-01

    We describe and evaluate various implicit and semi-implicit time integration schemes applied to the numerical simulation of hydrodynamical and magnetohydrodynamical problems. The schemes were implemented recently in the software package Versatile Advection Code, which uses modern shock capturing

  19. An ancient neurotrophin receptor code; a single Runx/Cbfβ complex determines somatosensory neuron fate specification in zebrafish.

    Science.gov (United States)

    Gau, Philia; Curtright, Andrew; Condon, Logan; Raible, David W; Dhaka, Ajay

    2017-07-01

    In terrestrial vertebrates such as birds and mammals, neurotrophin receptor expression is considered fundamental for the specification of distinct somatosensory neuron types where TrkA, TrkB and TrkC specify nociceptors, mechanoceptors and proprioceptors/mechanoceptors, respectively. In turn, Runx transcription factors promote neuronal fate specification by regulating neurotrophin receptor and sensory receptor expression where Runx1 mediates TrkA+ nociceptor diversification while Runx3 promotes a TrkC+ proprioceptive/mechanoceptive fate. Here, we report in zebrafish larvae that orthologs of the neurotrophin receptors in contrast to terrestrial vertebrates mark overlapping and distinct subsets of nociceptors suggesting that TrkA, TrkB and TrkC do not intrinsically promote nociceptor, mechanoceptor and proprioceptor/mechanoceptor neuronal fates, respectively. While we find that zebrafish Runx3 regulates nociceptors in contrast to terrestrial vertebrates, it shares a conserved regulatory mechanism found in terrestrial vertebrate proprioceptors/mechanoceptors in which it promotes TrkC expression and suppresses TrkB expression. We find that Cbfβ, which enhances Runx protein stability and affinity for DNA, serves as an obligate cofactor for Runx in neuronal fate determination. High levels of Runx can compensate for the loss of Cbfβ, indicating that in this context Cbfβ serves solely as a signal amplifier of Runx activity. Our data suggests an alteration/expansion of the neurotrophin receptor code of sensory neurons between larval teleost fish and terrestrial vertebrates, while the essential roles of Runx/Cbfβ in sensory neuron cell fate determination while also expanded are conserved.

  20. 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)

  1. 21 CFR 866.5520 - Immunoglobulin G (Fab fragment specific) immunological test system.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Immunoglobulin G (Fab fragment specific... Test Systems § 866.5520 Immunoglobulin G (Fab fragment specific) immunological test system. (a) Identification. An immunoglobulin G (Fab fragment specific) immunological test system is a device that consists...

  2. 21 CFR 866.5540 - Immunoglobulin G (Fd fragment specific) immunological test system.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Immunoglobulin G (Fd fragment specific... Test Systems § 866.5540 Immunoglobulin G (Fd fragment specific) immunological test system. (a) Identification. An immunoglobulin G (Fd fragment specific) immunological test system is a device that consists of...

  3. 21 CFR 866.5530 - Immunoglobulin G (Fc fragment specific) immunological test system.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Immunoglobulin G (Fc fragment specific... Test Systems § 866.5530 Immunoglobulin G (Fc fragment specific) immunological test system. (a) Identification. An immunoglobulin G (Fc fragment specific) immunological test system is a device that consists of...

  4. Testing, licensing, and code requirements for seismic isolation systems (for nuclear power plants)

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1987-01-01

    The use of seismic isolation as an earthquake hazard mitigation strategy for nuclear reactor power plants is rapidly receiving interest throughout the world. Seismic isolation has already been used on at least two French PWR plants, was to have been used for plants to be built in Iran, and is under serious consideration for advanced LMR plants (in the US, UK, France, and Japan). In addition, there is a growing use of seismic isolation throughout the world for other critical facilities such as hospitals, emergency facilities, buildings with very high-cost equipment (e.g., computers) and as a strategy to reduce loss of life and expensive equipment in earthquakes. Such a design approach is in complete contrast to the conventional seismic design strategy in which the structure and components are provided with sufficient strength and ductility to resist the earthquake forces and to prevent structural collapses or failure. The use of seismic isolation for nuclear plants can, therefore, be expected to be a significant licensing issue. For isolation, the licensing process must shift away in large measure from the superstructure and concentrate on the behavior of the seismic isolation system. This paper is not intended to promote the advantages of seismic isolation system, but to explore in some detail those technical issues which must be satisfactorily addressed to achieve full licensability of the use of seismic isolation as a viable, attractive and economical alternative to current traditional design approaches. Special problems and topics associated with testing and codes and standards development are addressed. A positive program for approach or strategy to secure licensing is presented

  5. Testing, licensing, and code requirements for seismic isolation systems (for nuclear power plants)

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.

    1987-01-01

    The use of seismic isolation as an earthquake hazard mitigation strategy for nuclear reactor power plants is rapidly receiving interest throughout the world. Seismic isolation has already been used on at least two French PWR plants, was to have been used for plants to be built in Iran, and is under serious consideration for advanced LMR plants (in the US, UK, France, and Japan). In addition, there is a growing use of seismic isolation throughout the world for other critical facilities such as hospitals, emergency facilities, buildings with very high-cost equipment (e.g., computers) and as a strategy to reduce loss of life and expensive equipment in earthquakes. Such a design approach is in complete contrast to the conventional seismic design strategy in which the structure and components are provided with sufficient strength and ductility to resist the earthquake forces and to prevent structural collapses or failure. The use of seismic isolation for nuclear plants can, therefore, be expected to be a significant licensing issue. For isolation, the licensing process must shift away in large measure from the superstructure and concentrate on the behavior of the seismic isolation system. This paper is not intended to promote the advantages of seismic isolation system, but to explore in some detail those technical issues which must be satisfactorily addressed to achieve full licensability of the use of seismic isolation as a viable, attractive and economical alternative to current traditional design approaches. Special problems and topics associated with testing and codes and standards development are addressed. A positive program for approach or strategy to secure licensing is presented.

  6. Benchmark test of drift-kinetic and gyrokinetic codes through neoclassical transport simulations

    International Nuclear Information System (INIS)

    Satake, S.; Sugama, H.; Watanabe, T.-H.; Idomura, Yasuhiro

    2009-09-01

    Two simulation codes that solve the drift-kinetic or gyrokinetic equation in toroidal plasmas are benchmarked by comparing the simulation results of neoclassical transport. The two codes are the drift-kinetic δf Monte Carlo code (FORTEC-3D) and the gyrokinetic full- f Vlasov code (GT5D), both of which solve radially-global, five-dimensional kinetic equation with including the linear Fokker-Planck collision operator. In a tokamak configuration, neoclassical radial heat flux and the force balance relation, which relates the parallel mean flow with radial electric field and temperature gradient, are compared between these two codes, and their results are also compared with the local neoclassical transport theory. It is found that the simulation results of the two codes coincide very well in a wide rage of plasma collisionality parameter ν * = 0.01 - 10 and also agree with the theoretical estimations. The time evolution of radial electric field and particle flux, and the radial profile of the geodesic acoustic mode frequency also coincide very well. These facts guarantee the capability of GT5D to simulate plasma turbulence transport with including proper neoclassical effects of collisional diffusion and equilibrium radial electric field. (author)

  7. Development and construction of a specific chamber for phototoxicity test

    Energy Technology Data Exchange (ETDEWEB)

    Sufi, Bianca S.; Mathor, Monica B., E-mail: biancasufi@usp.br, E-mail: mathor@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Esteves-Pedro, Natalia M.; Kaneko, Telma Mary, E-mail: nataliamenves@yahoo.com.br, E-mail: tsakuda@usp.br [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Fac. de Ciencias Farmaceuticas; Lopes, Patricia, E-mail: patricia.lopes@unifesp.br [Universidade Federal de Sao Paulo (UNIFESP), Diadema, SP (Brazil)

    2013-07-01

    Phototoxicity corresponds to the acute toxic response induced after skin exposure 'in vivo' and 'ex vivo' to certain chemicals and subsequent exposure to irradiation. Phototoxicity 'in vitro' assay is determined by viability of fibroblasts BALB/c 3T3 exposed to chemicals in the presence and absence of light. Substances identified as phototoxic are susceptible to 'in vivo' phototoxicity (OECD 432, 2004). A chamber was developed and constructed according to the guidelines OECD Toxicity Guide - 432 and ®ECVAM DB-ALM: INVITTOX N. 78. The chamber was built in stainless steel frame, with UVA lamps and dark area for negative control. The tests to qualify the chamber were performed with Sodium Lauryl Sulfate, recommended by the guides aforementioned, as negative control; and Bergamot oil (Givaudan-Roche), as positive control. Bergamot, Citrus bergamia, has, as major component, Bergapten responsible for its photosensitive activity. Both samples were diluted in Phosphate Buffered Saline with concentrations between 0.005 and 0.1 mg/mL, which were calculated by the dilution factor 1.47. These tests were performed over fibroblast BALB/c 3T3 culture and submitted to phototoxicity assay with MTS dye, under spectrophotometric reading, which allows determining the Photo Irritation Factor (PIF), what suggests that a substance with a PIF<2 predicts no phototoxicity; PIF>2 and <5 provides likely phototoxicity and PIF>5 provides phototoxicity. Sodium Lauryl Sulfate presented a PIF=1, being in accordance with the OECD. Bergamot oil has shown to be likely phototoxic with a PIF=2,475. These results provide that the chamber is qualified to be used to perform phototoxicity tests with assurance and security. (author)

  8. Development and construction of a specific chamber for phototoxicity test

    International Nuclear Information System (INIS)

    Sufi, Bianca S.; Mathor, Monica B.; Esteves-Pedro, Natalia M.; Kaneko, Telma Mary

    2013-01-01

    Phototoxicity corresponds to the acute toxic response induced after skin exposure 'in vivo' and 'ex vivo' to certain chemicals and subsequent exposure to irradiation. Phototoxicity 'in vitro' assay is determined by viability of fibroblasts BALB/c 3T3 exposed to chemicals in the presence and absence of light. Substances identified as phototoxic are susceptible to 'in vivo' phototoxicity (OECD 432, 2004). A chamber was developed and constructed according to the guidelines OECD Toxicity Guide - 432 and ®ECVAM DB-ALM: INVITTOX N. 78. The chamber was built in stainless steel frame, with UVA lamps and dark area for negative control. The tests to qualify the chamber were performed with Sodium Lauryl Sulfate, recommended by the guides aforementioned, as negative control; and Bergamot oil (Givaudan-Roche), as positive control. Bergamot, Citrus bergamia, has, as major component, Bergapten responsible for its photosensitive activity. Both samples were diluted in Phosphate Buffered Saline with concentrations between 0.005 and 0.1 mg/mL, which were calculated by the dilution factor 1.47. These tests were performed over fibroblast BALB/c 3T3 culture and submitted to phototoxicity assay with MTS dye, under spectrophotometric reading, which allows determining the Photo Irritation Factor (PIF), what suggests that a substance with a PIF 2 and 5 provides phototoxicity. Sodium Lauryl Sulfate presented a PIF=1, being in accordance with the OECD. Bergamot oil has shown to be likely phototoxic with a PIF=2,475. These results provide that the chamber is qualified to be used to perform phototoxicity tests with assurance and security. (author)

  9. 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)

  10. 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)

  11. 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.

  12. Error-correction coding

    Science.gov (United States)

    Hinds, Erold W. (Principal Investigator)

    1996-01-01

    This report describes the progress made towards the completion of a specific task on error-correcting coding. The proposed research consisted of investigating the use of modulation block codes as the inner code of a concatenated coding system in order to improve the overall space link communications performance. The study proposed to identify and analyze candidate codes that will complement the performance of the overall coding system which uses the interleaved RS (255,223) code as the outer code.

  13. Comparison of Crack Growth Test Results at Elevated Temperature and Design Code Material Properties for Grade 91 Steel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon; Kim, Woo-Gon; Kim, Nak-Hyun [Korea Atomic Energy Reserach Institute, Daejeon (Korea, Republic of)

    2015-01-15

    The material properties of crack growth models at an elevated temperature were derived from the results of numerous crack growth tests for Mod.9Cr-1Mo (ASME Grade 91) steel specimens under fatigue loading and creep loading at an elevated temperature. These crack growth models were needed for defect assessment under creep-fatigue loading. The mathematical crack growth rate models for fatigue crack growth (FCG) and creep crack growth (CCG) were determined based on the test results, and the models were compared with those of the French design code RCCMRx to investigate the conservatism of the code. The French design code RCC-MRx provides an FCG model and a CCG model for Grade 91 steel in Section III Tome 6. It was shown that the FCG model of RCC-MRx is conservative, while the CCG model is non-conservative compared with the present test data. Thus, it was shown that further validation of the property was required. Mechanical strength tests and creep tests were also conducted, and the test results were compared with those of RCC-MRx.

  14. Agile deployment and code coverage testing metrics of the boot software on-board Solar Orbiter's Energetic Particle Detector

    Science.gov (United States)

    Parra, Pablo; da Silva, Antonio; Polo, Óscar R.; Sánchez, Sebastián

    2018-02-01

    In this day and age, successful embedded critical software needs agile and continuous development and testing procedures. This paper presents the overall testing and code coverage metrics obtained during the unit testing procedure carried out to verify the correctness of the boot software that will run in the Instrument Control Unit (ICU) of the Energetic Particle Detector (EPD) on-board Solar Orbiter. The ICU boot software is a critical part of the project so its verification should be addressed at an early development stage, so any test case missed in this process may affect the quality of the overall on-board software. According to the European Cooperation for Space Standardization ESA standards, testing this kind of critical software must cover 100% of the source code statement and decision paths. This leads to the complete testing of fault tolerance and recovery mechanisms that have to resolve every possible memory corruption or communication error brought about by the space environment. The introduced procedure enables fault injection from the beginning of the development process and enables to fulfill the exigent code coverage demands on the boot software.

  15. SU-E-T-180: Fano Cavity Test of Proton Transport in Monte Carlo Codes Running On GPU and Xeon Phi

    International Nuclear Information System (INIS)

    Sterpin, E; Sorriaux, J; Souris, K; Lee, J; Vynckier, S; Schuemann, J; Paganetti, H; Jia, X; Jiang, S

    2014-01-01

    Purpose: In proton dose calculation, clinically compatible speeds are now achieved with Monte Carlo codes (MC) that combine 1) adequate simplifications in the physics of transport and 2) the use of hardware architectures enabling massive parallel computing (like GPUs). However, the uncertainties related to the transport algorithms used in these codes must be kept minimal. Such algorithms can be checked with the so-called “Fano cavity test”. We implemented the test in two codes that run on specific hardware: gPMC on an nVidia GPU and MCsquare on an Intel Xeon Phi (60 cores). Methods: gPMC and MCsquare are designed for transporting protons in CT geometries. Both codes use the method of fictitious interaction to sample the step-length for each transport step. The considered geometry is a water cavity (2×2×0.2 cm 3 , 0.001 g/cm 3 ) in a 10×10×50 cm 3 water phantom (1 g/cm 3 ). CPE in the cavity is established by generating protons over the phantom volume with a uniform momentum (energy E) and a uniform intensity per unit mass I. Assuming no nuclear reactions and no generation of other secondaries, the computed cavity dose should equal IE, according to Fano's theorem. Both codes were tested for initial proton energies of 50, 100, and 200 MeV. Results: For all energies, gPMC and MCsquare are within 0.3 and 0.2 % of the theoretical value IE, respectively (0.1% standard deviation). Single-precision computations (instead of double) increased the error by about 0.1% in MCsquare. Conclusion: Despite the simplifications in the physics of transport, both gPMC and MCsquare successfully pass the Fano test. This ensures optimal accuracy of the codes for clinical applications within the uncertainties on the underlying physical models. It also opens the path to other applications of these codes, like the simulation of ion chamber response

  16. SU-E-T-180: Fano Cavity Test of Proton Transport in Monte Carlo Codes Running On GPU and Xeon Phi

    Energy Technology Data Exchange (ETDEWEB)

    Sterpin, E; Sorriaux, J; Souris, K; Lee, J; Vynckier, S [Universite catholique de Louvain, Brussels, Brussels (Belgium); Schuemann, J; Paganetti, H [Massachusetts General Hospital, Boston, MA (United States); Jia, X; Jiang, S [The University of Texas Southwestern Medical Ctr, Dallas, TX (United States)

    2014-06-01

    Purpose: In proton dose calculation, clinically compatible speeds are now achieved with Monte Carlo codes (MC) that combine 1) adequate simplifications in the physics of transport and 2) the use of hardware architectures enabling massive parallel computing (like GPUs). However, the uncertainties related to the transport algorithms used in these codes must be kept minimal. Such algorithms can be checked with the so-called “Fano cavity test”. We implemented the test in two codes that run on specific hardware: gPMC on an nVidia GPU and MCsquare on an Intel Xeon Phi (60 cores). Methods: gPMC and MCsquare are designed for transporting protons in CT geometries. Both codes use the method of fictitious interaction to sample the step-length for each transport step. The considered geometry is a water cavity (2×2×0.2 cm{sup 3}, 0.001 g/cm{sup 3}) in a 10×10×50 cm{sup 3} water phantom (1 g/cm{sup 3}). CPE in the cavity is established by generating protons over the phantom volume with a uniform momentum (energy E) and a uniform intensity per unit mass I. Assuming no nuclear reactions and no generation of other secondaries, the computed cavity dose should equal IE, according to Fano's theorem. Both codes were tested for initial proton energies of 50, 100, and 200 MeV. Results: For all energies, gPMC and MCsquare are within 0.3 and 0.2 % of the theoretical value IE, respectively (0.1% standard deviation). Single-precision computations (instead of double) increased the error by about 0.1% in MCsquare. Conclusion: Despite the simplifications in the physics of transport, both gPMC and MCsquare successfully pass the Fano test. This ensures optimal accuracy of the codes for clinical applications within the uncertainties on the underlying physical models. It also opens the path to other applications of these codes, like the simulation of ion chamber response.

  17. Insertion of control systems models in the Almod 3 computer code for the simulation of Angra I reactor start-up tests

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1981-09-01

    The Almod 3 computer code was modified, aiming at the simulation of Angra I nuclear power plant behavior during some reactor start-up tests. The results obtained with the modified computer code (Almod 3W) are compared with those obtained with the Retran computer code. (E.G.) [pt

  18. 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

  19. 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.

  20. 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

  1. 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)

  2. Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

    International Nuclear Information System (INIS)

    Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras

    2003-01-01

    This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

  3. 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)

  4. 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.

  5. Flica: a code for the thermodynamic study of a reactor or a test loop

    International Nuclear Information System (INIS)

    Fajeau, M.

    1969-01-01

    This code handles the thermal problems of water loops or reactor cores under the following conditions: High or low pressure, steady state or transient behavior, one or two phases - Three-dimensional thermodynamic study of the flow in cylindrical geometry - Unidimensional study of heat transfer in heating elements - Neutronic studies can be coupled and a schematic representation of the safety rod behavior is given. The number of cells described in a flow cross-section is presently less than 20. This code is the logical following of FLID and CACTUS of which it constitutes a synthesis. (author) [fr

  6. Validating a dance-specific screening test for balance: preliminary results from multisite testing.

    Science.gov (United States)

    Batson, Glenna

    2010-09-01

    Few dance-specific screening tools adequately capture balance. The aim of this study was to administer and modify the Star Excursion Balance Test (oSEBT) to examine its utility as a balance screen for dancers. The oSEBT involves standing on one leg while lightly targeting with the opposite foot to the farthest distance along eight spokes of a star-shaped grid. This task simulates dance in the spatial pattern and movement quality of the gesturing limb. The oSEBT was validated for distance on athletes with history of ankle sprain. Thirty-three dancers (age 20.1 +/- 1.4 yrs) participated from two contemporary dance conservatories (UK and US), with or without a history of lower extremity injury. Dancers were verbally instructed (without physical demonstration) to execute the oSEBT and four modifications (mSEBT): timed (speed), timed with cognitive interference (answering questions aloud), and sensory disadvantaging (foam mat). Stepping strategies were tracked and performance strategies video-recorded. Unlike the oSEBT results, distances reached were not significant statistically (p = 0.05) or descriptively (i.e., shorter) for either group. Performance styles varied widely, despite sample homogeneity and instructions to control for strategy. Descriptive analysis of mSEBT showed an increased number of near-falls and decreased timing on the injured limb. Dancers appeared to employ variable strategies to keep balance during this test. Quantitative analysis is warranted to define balance strategies for further validation of SEBT modifications to determine its utility as a balance screening tool.

  7. Modelling the attenuation in the ATHENA finite elements code for the ultrasonic testing of austenitic stainless steel welds.

    Science.gov (United States)

    Chassignole, B; Duwig, V; Ploix, M-A; Guy, P; El Guerjouma, R

    2009-12-01

    Multipass welds made in austenitic stainless steel, in the primary circuit of nuclear power plants with pressurized water reactors, are characterized by an anisotropic and heterogeneous structure that disturbs the ultrasonic propagation and makes ultrasonic non-destructive testing difficult. The ATHENA 2D finite element simulation code was developed to help understand the various physical phenomena at play. In this paper, we shall describe the attenuation model implemented in this code to give an account of wave scattering phenomenon through polycrystalline materials. This model is in particular based on the optimization of two tensors that characterize this material on the basis of experimental values of ultrasonic velocities attenuation coefficients. Three experimental configurations, two of which are representative of the industrial welds assessment case, are studied in view of validating the model through comparison with the simulation results. We shall thus provide a quantitative proof that taking into account the attenuation in the ATHENA code dramatically improves the results in terms of the amplitude of the echoes. The association of the code and detailed characterization of a weld's structure constitutes a remarkable breakthrough in the interpretation of the ultrasonic testing on this type of component.

  8. 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)

  9. 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

  10. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

    2013-12-15

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

  11. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

    2013-01-01

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

  12. Inspection, testing, and operating requiremens for the packaging and shipping of uranium trioxide in 55-gallon Department of Transportation (DOT) Specification 6M shipping packagings

    International Nuclear Information System (INIS)

    Toomer, D.V.

    1991-06-01

    This document identifies the inspection, testing and operating requirements for the packaging, loading, and shipping of uranium trioxide (UO 3 ) in 55-gallon DOT Specification 6M shipping packagings from the Idaho Chemical Processing Plant (ICPP). Compliance with this document assures established controls for the purchasing, packaging, loading, and shipping of DOT Specification 6M shipping packagings are maintained in strict accordance with applicable Code of Federal Regulations (CFRs) and Department of Energy (DOE) Orders. 7 refs., 3 figs., 1 tab

  13. Pain elicited by the Cold Pressor Test: A gender-comparative FACS coding study of spontaneous, faked and inhibited expressions.

    OpenAIRE

    Gil, Luisa; De Sousa, Cristina; Baunninger-Huber, Eva; Schiestl, Cathrin; Toussaint, Kyra; Gruber, Verena; Oliveira, Armando Monica; Duarte, Ana Catarina

    2012-01-01

    Evolutionary theories of pain have conjectured a better ability of males to control their facial expressions of pain, and of females to express and communicate emotions through the face. The present study involved 24 participants (12 men; 12 women). Pain was induced via the Cold Pressor Test (CPT), and three expressive contexts (spontaneous, faked an inhibited) were created through instructions. Elicited pain expressions were FACS coded and frequency, indices were derived for the observed Act...

  14. MACRO1: a code to test a methodology for analyzing nuclear-waste management systems

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1979-01-01

    The code is primarily a manager of probabilistic data and deterministic mathematical models. The user determines the desired aggregation of the available models into a composite model of a physical system. MACRO1 then propagates the finite probability distributions of the inputs to the model to finite probability distributions over the outputs. MACRO1 has been applied to a sample analysis of a nuclear-waste repository, and its results compared satisfactorily with previously obtained Monte Carlo statistics

  15. Development of new test procedures for measuring fine and coarse aggregates specific gravity.

    Science.gov (United States)

    2009-09-01

    The objective of the research is to develop and evaluate new test methods at determining the specific gravity and absorption of both fine and coarse aggregates. Current methods at determining the specific gravity and absorption of fine and coarse agg...

  16. Development of explicit solution scheme for the MATRA-LMR code and test calculation

    International Nuclear Information System (INIS)

    Jeong, H. Y.; Ha, K. S.; Chang, W. P.; Kwon, Y. M.; Jeong, K. S.

    2003-01-01

    The local blockage in a subassembly of a liquid metal reactor is of particular importance because local sodium boiling could occur at the downstream of the blockage and integrity of the fuel clad could be threatened. The explicit solution scheme of MATRA-LMR code is developed to analyze the flow blockage in a subassembly of a liquid metal cooled reactor. In the present study, the capability of the code is extended to the analysis of complete blockage of one or more subchannels. The results of the developed solution scheme shows very good agreement with the results obtained from the implicit scheme for the experiments of flow channel without any blockage. The applicability of the code is also evaluated for two typical experiments in a blocked channel. Through the sensitivity study, it is shown that the explicit scheme of MATRA-LMR predicts the flow and temperature profile after blockage reasonably if the effect of wire is suitably modeled. The simple assumption in wire-forcing function is effective for the un-blocked case or for the case of blockage with lower velocity. A different type of wire-forcing function describing the velocity reduction after blockage or an accurate distributed resistance model is required for more improved predictions

  17. Pre-Analysis for Safety-Related Verification Test Using TASS/SMR Code

    Energy Technology Data Exchange (ETDEWEB)

    Ra, I. S.; Kim, H. J.; Jeon, G. H. [ACTS Ltd., Daejeon (Korea, Republic of)

    2010-01-15

    General trends of TASS/SMR simulation were similar to those in both ORNF test and BENNETT test conducted to verify core heat transfer model in TASS/SMR. In high mass flux, however, a CHF location in the analytical result of TASS/SMR was greatly deviated from BENNETT test result. TASS/SMR gave better results in heterogeneous option that in homogeneous option in both KIT test, which was a steady state test with an inlet flow, and GE-LEVEL Swell test, which a transient test without an inlet flow. TASS/SMR simulation for SMD Long and Short test gave a good agreement with the test results in showing a reasonable predictability of critical flow model. But, in the case of Marviken test, the analytical result was not similar to the test result after the timing of vapor generation

  18. Specific diversity and morphological indices of muriform rodents in some areas of Semipalatinsk test range zone

    International Nuclear Information System (INIS)

    Magda, I.N.; Chernykh, A.B.; Morozov, A.E.; Bushneva, I.A.; Ponyavkina, A.G.

    2002-01-01

    There were presented the results of the preliminary estimation of comparative specific diversity and morphological indices of muriform rodents inhabiting separate areas of the Semipalatinsk test site. (author)

  19. Testing constancy of unconditional variance in volatility models by misspecification and specification tests

    DEFF Research Database (Denmark)

    Silvennoinen, Annastiina; Terasvirta, Timo

    The topic of this paper is testing the hypothesis of constant unconditional variance in GARCH models against the alternative that the unconditional variance changes deterministically over time. Tests of this hypothesis have previously been performed as misspecification tests after fitting a GARCH...... models. An application to exchange rate returns is included....

  20. Soil-Geosynthetic Interaction Test to Develop Specifications for Geosynthetic-Stabilized Roadways

    Science.gov (United States)

    2018-05-01

    soil-geosynthetic composite (KSGC) for a wide range of geosynthetics. The tests were conducted after establishment of test configurations that were found suitable for specification of geosynthetic-stabilized base roadways. Field performance of experi...

  1. Test of user- and system programs coded in real time languages - requirements on program language and testing tool

    International Nuclear Information System (INIS)

    Hertlin, J.; Mackert, M.

    1979-01-01

    In the present paper the functions are presented, which should be part of a test system for user programs in a higher treat time programming language, taking into account time sequences and competitive processes. As can be shown by the problem of testing, use of higher level real time programming languages renders the task of program development essentially easier, however performance of test procedures without appropriate test systems is very difficult. After the presentation of notions and methods for the testing of programs, general requirements on testing tools are described and the test system functions for a program test, beeing uncritical with respect to time, are placed together. Thereby, for every individual function, the interface between the test system, the program under test, and the residual program-generation system (compiler, binder, operating system, delay-time system, and loader) is given too. For the time-critical test, a series of desirable functions are described, which can be implemented with acceptable expense. (orig.) [de

  2. The end of the road for prostate specific antigen testing? | Nna ...

    African Journals Online (AJOL)

    Many candidate biomarkers for diagnosis of prostate cancer have been investigated, but prostate‑specific antigen (PSA) testing remains the frontline test for both mass screening and individual clinical testing. Although the PSA test is cost‑effective, analytically reliable, and flexibly high throughput, it has a very weak ...

  3. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  4. Testing a solar coronal magnetic field extrapolation code with the Titov–Démoulin magnetic flux rope model

    International Nuclear Information System (INIS)

    Jiang, Chao-Wei; Feng, Xue-Shang

    2016-01-01

    In the solar corona, the magnetic flux rope is believed to be a fundamental structure that accounts for magnetic free energy storage and solar eruptions. Up to the present, the extrapolation of the magnetic field from boundary data has been the primary way to obtain fully three-dimensional magnetic information about the corona. As a result, the ability to reliably recover the coronal magnetic flux rope is important for coronal field extrapolation. In this paper, our coronal field extrapolation code is examined with an analytical magnetic flux rope model proposed by Titov and Démoulin, which consists of a bipolar magnetic configuration holding a semi-circular line-tied flux rope in force-free equilibrium. By only using the vector field at the bottom boundary as input, we test our code with the model in a representative range of parameter space and find that the model field can be reconstructed with high accuracy. In particular, the magnetic topological interfaces formed between the flux rope and the surrounding arcade, i.e., the “hyperbolic flux tube” and “bald patch separatrix surface,” are also reliably reproduced. By this test, we demonstrate that our CESE–MHD–NLFFF code can be applied to recovering the magnetic flux rope in the solar corona as long as the vector magnetogram satisfies the force-free constraints. (paper)

  5. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Seniuk, P.J.

    1996-12-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

  6. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    International Nuclear Information System (INIS)

    Seniuk, P.J.

    1996-01-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, open-quotes Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plantsclose quotes. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O ampersand M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O ampersand M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion

  7. Severe accident tests and development of domestic severe accident system codes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  8. Severe accident tests and development of domestic severe accident system codes

    International Nuclear Information System (INIS)

    2013-01-01

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  9. Use of the FLUKA Monte Carlo code for 3D patient-specific dosimetry on PET-CT and SPECT-CT images*

    Science.gov (United States)

    Botta, F; Mairani, A; Hobbs, R F; Vergara Gil, A; Pacilio, M; Parodi, K; Cremonesi, M; Coca Pérez, M A; Di Dia, A; Ferrari, M; Guerriero, F; Battistoni, G; Pedroli, G; Paganelli, G; Torres Aroche, L A; Sgouros, G

    2014-01-01

    Patient-specific absorbed dose calculation for nuclear medicine therapy is a topic of increasing interest. 3D dosimetry at the voxel level is one of the major improvements for the development of more accurate calculation techniques, as compared to the standard dosimetry at the organ level. This study aims to use the FLUKA Monte Carlo code to perform patient-specific 3D dosimetry through direct Monte Carlo simulation on PET-CT and SPECT-CT images. To this aim, dedicated routines were developed in the FLUKA environment. Two sets of simulations were performed on model and phantom images. Firstly, the correct handling of PET and SPECT images was tested under the assumption of homogeneous water medium by comparing FLUKA results with those obtained with the voxel kernel convolution method and with other Monte Carlo-based tools developed to the same purpose (the EGS-based 3D-RD software and the MCNP5-based MCID). Afterwards, the correct integration of the PET/SPECT and CT information was tested, performing direct simulations on PET/CT images for both homogeneous (water) and non-homogeneous (water with air, lung and bone inserts) phantoms. Comparison was performed with the other Monte Carlo tools performing direct simulation as well. The absorbed dose maps were compared at the voxel level. In the case of homogeneous water, by simulating 108 primary particles a 2% average difference with respect to the kernel convolution method was achieved; such difference was lower than the statistical uncertainty affecting the FLUKA results. The agreement with the other tools was within 3–4%, partially ascribable to the differences among the simulation algorithms. Including the CT-based density map, the average difference was always within 4% irrespective of the medium (water, air, bone), except for a maximum 6% value when comparing FLUKA and 3D-RD in air. The results confirmed that the routines were properly developed, opening the way for the use of FLUKA for patient-specific, image

  10. A test of the IAEA code of practice for absorbed dose determination in photon and electron beams

    International Nuclear Information System (INIS)

    Leitner, A.; Tiefenboeck, W.; Witzani, J.; Strachotinsky, C.

    1990-12-01

    The IAEA Code of Practice TRS 277 gives recommendations for absorbed dose determination in high energy photon and electron beams based on the use of ionisation chambers calibrated in terms of exposure or air kerma. The scope of the present work was to test the Code for 60 Co gamma radiation and for several radiation qualities at four different types of electron accelerators and to compare the ionisation chamber dosimetry with ferrous sulphate dosimetry. The results show agreement between the two methods within about one per cent for all the investigated qualities. In addition the response of the TLD capsules of the IAEA/WHO TL dosimetry service has been determined. (Authors) 5 refs., 9 tabs., 3 figs

  11. Project W-314 specific test and evaluation plan for 241-AY-02A pump pit upgrade

    International Nuclear Information System (INIS)

    Hays, W.H.

    1998-01-01

    This Specific Test and Evaluation Plan (STEP) defines the test and evaluation activities encompassing the upgrade of the 241-AY-02A Pump Pit for the W-314 Project. The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AY-02A Pump Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation Plan (TEP)

  12. Project W-314 specific test and evaluation plan for 241-AY-01A pump pit upgrade

    International Nuclear Information System (INIS)

    Hays, W.H.

    1998-01-01

    This Specific Test and Evaluation Plan (STEP) defines the test and evaluation activities encompassing the upgrade of the 241-AY-0IA Pump Pit for the W-314 Project. The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AY-01A Pump Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation Plan (TEP)

  13. Empirical usability testing in a component-based environment : improving test efficiency with component-specific usability measures

    NARCIS (Netherlands)

    Brinkman, W.P.; Haakma, R.; Bouwhuis, D.G.; Bastide, R.; Palanque, P.; Roth, J.

    2005-01-01

    This paper addresses the issue of usability testing in a component-based software engineering environment, specifically measuring the usability of different versions of a component in a more powerful manner than other, more holistic, usability methods. Three component-specific usability measures are

  14. 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)

  15. Test- and behavior-specific genetic factors affect WKY hypoactivity in tests of emotionality.

    Science.gov (United States)

    Baum, Amber E; Solberg, Leah C; Churchill, Gary A; Ahmadiyeh, Nasim; Takahashi, Joseph S; Redei, Eva E

    2006-05-15

    Inbred Wistar-Kyoto rats consistently display hypoactivity in tests of emotional behavior. We used them to test the hypothesis that the genetic factors underlying the behavioral decision-making process will vary in different environmental contexts. The contexts used were the open-field test (OFT), a novel environment with no explicit threats present, and the defensive-burying test (DB), a habituated environment into which a threat has been introduced. Rearing, a voluntary behavior was measured in both tests, and our study was the first to look for genetic loci affecting grooming, a relatively automatic, stress-responsive stereotyped behavior. Quantitative trait locus analysis was performed on a population of 486 F2 animals bred from reciprocal inter-crosses. The genetic architectures of DB and OFT rearing, and of DB and OFT grooming, were compared. There were no common loci affecting grooming behavior in both tests. These different contexts produced the stereotyped behavior via different pathways, and genetic factors seem to influence the decision-making pathways and not the expression of the behavior. Three loci were found that affected rearing behavior in both tests. However, in both contexts, other loci had greater effects on the behavior. Our results imply that environmental context's effects on decision-making vary depending on the category of behavior.

  16. Enclosure environment characterization testing for the base line validation of computer fire simulation codes

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1987-03-01

    This report describes a series of fire tests conducted under the direction of Sandia National Laboratories for the US Nuclear Regulatory Commission. The primary purpose of these tests was to provide data against which to validate computer fire environment simulation models to be used in the analysis of nuclear power plant enclosure fire situations. Examples of the data gathered during three of the tests are presented, though the primary objective of this report is to provide a timely description of the test effort itself. These tests were conducted in an enclosure measuring 60x40x20 feet constructed at the Factory Mutual Research Corporation fires test facility in Rhode Island. All of the tests utilized forced ventilation conditions. The ventilation system was designed to simulate typical nuclear power plant installation practices and ventilation rates. A total of 22 tests using simple gas burner, heptane pool, methanol pool, and PMMA solid fires was conducted. Four of these tests were conducted with a full-scale control room mockup in place. Parameters varied during testing were fire intensity, enclosure ventilation rate, and fire location. Data gathered include air temperatures, air velocities, radiative and convective heat flux levels, optical smoke densities, inner and outer enclosure surface temperatures, enclosure surface heat flux levels, and gas concentrations within the enclosure in the exhaust stream

  17. Relationships between the handball-specific complex test, non-specific field tests and the match performance score in elite professional handball players.

    Science.gov (United States)

    Hermassi, Souhail; Chelly, Mohamed-Souhaiel; Wollny, Rainer; Hoffmeyer, Birgit; Fieseler, Georg; Schulze, Stephan; Irlenbusch, Lars; Delank, Karl-Stefan; Shephard, Roy J; Bartels, Thomas; Schwesig, René

    2018-06-01

    This study assessed the validity of the handball-specific complex test (HBCT) and two non-specific field tests in professional elite handball athletes, using the match performance score (MPS) as the gold standard of performance. Thirteen elite male handball players (age: 27.4±4.8 years; premier German league) performed the HBCT, the Yo-Yo Intermittent Recovery (YYIR) test and a repeated shuttle sprint ability (RSA) test at the beginning of pre-season training. The RSA results were evaluated in terms of best time, total time, and fatigue decrement. Heart rates (HR) were assessed at selected times throughout all tests; the recovery HR was measured immediately post-test and 10 minutes later. The match performance score was based on various handball specific parameters (e.g., field goals, assists, steals, blocks, and technical mistakes) as seen during all matches of the immediately subsequent season (2015/2016). The parameters of run 1, run 2, and HR recovery at minutes 6 and 10 of the RSA test all showed a variance of more than 10% (range: 11-15%). However, the variance of scores for the YYIR test was much smaller (range: 1-7%). The resting HR (r2=0.18), HR recovery at minute 10 (r2=0.10), lactate concentration at rest (r2=0.17), recovery of heart rate from 0 to 10 minutes (r2=0.15), and velocity of second throw at first trial (r2=0.37) were the most valid HBCT parameters. Much effort is necessary to assess MPS and to develop valid tests. Speed and the rate of functional recovery seem the best predictors of competitive performance for elite handball players.

  18. STUDY ON MAXIMUM SPECIFIC SLUDGE ACIVITY OF DIFFERENT ANAEROBIC GRANULAR SLUDGE BY BATCH TESTS

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The maximum specific sludge activity of granular sludge from large-scale UASB, IC and Biobed anaerobic reactors were investigated by batch tests. The limitation factors related to maximum specific sludge activity (diffusion, substrate sort, substrate concentration and granular size) were studied. The general principle and procedure for the precise measurement of maximum specific sludge activity were suggested. The potential capacity of loading rate of the IC and Biobed anaerobic reactors were analyzed and compared by use of the batch tests results.

  19. Analysis of recent post irradiation tests by Japanese and French burnup analysis code systems

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko; Hiraizumi, Hiroaki; Youinou, Gilles

    2002-01-01

    Benchmark problem based on Japanese Post Irradiation Experiment (PIE) data was analyzed by Japanese burnup analysis code and French one under the cooperative research program between the Japanese University Association (JUA) in Japan and Commissariat a l'Enegie Atomique (CEA) in France. Significant discrepancies over 10% were found between the Japanese and French results for 238 Pu, 243 Am, 244 Cm, 125 Sb, 154 Eu, 134 Cs and 144 Ce. It is supposed that the difference of C/E for 243 Am and 244 Cm between Japanese results and French ones is due to the (n,gamma) reaction of 242m Am. For 125 Sb and 154 Eu, the C/E values are improved by using new cross section and fission yield libraries. (author)

  20. Drilling and testing specifications for RRL-6, RRL-14, RRL-15 and DC-3

    International Nuclear Information System (INIS)

    Moak, D.J.

    1982-07-01

    RRL-6, RRL-14, RRL-15, and DC-3 will provide data for characterization of the stratigraphy and intraflow structures in the Reference Repository Location. This test specification includes details for the drilling and testing of the boreholes. It includes the predicted stratigraphy, the drilling requirements, description of tests to be conducted, intervals selected for hydrologic testing and a schedule of the drilling and testing activities. 14 refs., 8 figs., 12 tabs

  1. 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.

  2. 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

  3. HPV specific testing: a requirement for oropharyngeal squamous cell carcinoma patients.

    Science.gov (United States)

    Robinson, Max; Schache, Andrew; Sloan, Philip; Thavaraj, Selvam

    2012-07-01

    Human papillomavirus (HPV) testing is now recommended as part of the work up for patients with oropharyngeal squamous cell carcinoma (OPSCC) and those patients with cervical lymph node metastasis of unknown origin. The laboratory testing strategy should accurately assess the presence or absence of oncogenic HPV infection in routinely collected tumour samples that are subject to standard fixation protocols, alcohol-fixed cytological preparations and formalin-fixed tissue samples. The HPV status should correlate with biologically relevant outcome measures such as overall, disease-specific and disease-free survival. Whilst increased expression of p16 by immunohistochemistry is considered to be a surrogate marker of oncogenic HPV infection and is a validated independent prognostic biomarker, only HPV specific tests provide definitive evidence of the aetiological agent. We provide an overview of HPV testing in OPSCC, justifying the use of HPV specific tests. We examine the analytical accuracy of HPV specific tests against the 'reference' test--high risk HPV mRNA in fresh tissue--and contrast this with the performance of p16 immunohistochemistry as a stand alone test. We highlight the added value of HPV specific tests in prognostication, clinical trial design, and population-based disease surveillance. We consider that HPV specific testing is the starting point for developing increasingly informative biomarker panels in the context of 'stratified medicine'. We briefly frame test information in the context of disclosure of HPV status to patients. We conclude that only a testing strategy that includes HPV specific tests can deliver more effective care for patients with OPSCC. The international head and neck oncology community should work together to clearly define the minimum requirements for assigning a diagnosis of HPV-related OPSCC in order to ensure consistent reporting of this emerging and increasingly prevalent disease.

  4. 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)

  5. 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.

  6. 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

  7. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    International Nuclear Information System (INIS)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

    2016-01-01

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

  8. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  9. Validation of the Serpent 2-DYNSUB code sequence using the Special Power Excursion Reactor Test III (SPERT III)

    International Nuclear Information System (INIS)

    Knebel, Miriam; Mercatali, Luigi; Sanchez, Victor; Stieglitz, Robert; Macian-Juan, Rafael

    2016-01-01

    Highlights: • Full few-group cross section tables created by Monte Carlo lattice code Serpent 2. • Serpent 2 group constant methodology verified for HFP static and transient cases. • Serpent 2-DYNSUB tool chainvalidated using SPERT III REA experiments. • Serpent 2-DYNSUB tool chain suitable to model RIAs in PWRs. - Abstract: The Special Power Excursion Reactor Test III (SPERT III) is studied using the Serpent 2-DYNSUB code sequence in order to validate it for modeling reactivity insertion accidents (RIA) in PWRs. The SPERT III E-core was a thermal research reactor constructed to analyze reactor dynamics. Its configuration resembles a commercial PWR on terms of fuel type, choice of moderator, coolant flow and system pressure. The initial conditions of the rod ejection accident experiments (REA) performed cover cold startup, hot startup, hot standby and operating power scenarios. Eight of these experiments were analyzed in detail. Firstly, multi-dimensional nodal diffusion cross section tables were created for the three-dimensional reactor simulator DYNSUB employing the Monte Carlo neutron transport code Serpent 2. In a second step, DYNSUB stationary simulations were compared to Monte Carlo reference three-dimensional full scale solutions obtained with Serpent 2 (cold startup conditions) and Serpent 2/SUBCHANFLOW (operating power conditions) with a good agreement being observed. The latter tool is an internal coupling of Serpent 2 and the sub-channel thermal-hydraulics code SUBCHANFLOW. Finally, DYNSUB was utilized to study the eight selected transient experiments. Results were found to match measurements well. As the selected experiments cover much of the possible transient (delayed super-critical, prompt super-critical and super-prompt critical excursion) and initial conditions (cold and hot as well as zero, little and full power reactor states) one expects in commercial PWRs, the obtained results give confidence that the Serpent 2-DYNSUB tool chain is

  10. Project W-314 specific test and evaluation plan for 241-AN-A valve pit

    International Nuclear Information System (INIS)

    Hays, W.H.

    1997-01-01

    The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AN-A Valve Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a ''lower tier'' document based on the W-314 Test and Evaluation Plan (TEP) This STEP encompasses all testing activities required to demonstrate compliance to the project design criteria as it relates to the modifications of the AN-A valve pit. The Project Design Specifications (PDS) identify the specific testing activities required for the Project. Testing includes Validations and Verifications (e.g., Commercial Grade Item Dedication activities), Factory Acceptance Tests (FATs), installation tests and inspections, Construction Acceptance Tests (CATs), Acceptance Test Procedures (ATPs), Pre-Operational Test Procedures (POTPs), and Operational Test Procedures (OTPs). It should be noted that POTPs are not required for testing of the modifications to the 241-AN-A Valve Pit. The STEP will be utilized in conjunction with the TEP for verification and validation

  11. Simulation of VVER MCCI reactor test case with ASTEC V2/MEDICIS computer code

    International Nuclear Information System (INIS)

    Stefanova, A.; Grudev, P.; Gencheva, R.

    2011-01-01

    This paper presents an application of the ASTEC v2, module MEDICIS for simulation of VVER Molten core concrete interaction test (MCCI) case without water injection. The main purpose of performed calculation is verification and improvement of module MEDICIS/ASTECv2 for better simulation of core concrete interaction processes. The VVER-1000 reference nuclear power plant was chosen as SARNET2 benchmark MCCI test-case. The initial conditions for MCCI test are taken after SBO scenario calculated with ASTEC version 1.3R2 by INRNE. (authors)

  12. 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

  13. Application of a path sensitizing method on automated generation of test specifications for control software

    International Nuclear Information System (INIS)

    Morimoto, Yuuichi; Fukuda, Mitsuko

    1995-01-01

    An automated generation method for test specifications has been developed for sequential control software in plant control equipment. Sequential control software can be represented as sequential circuits. The control software implemented in a control equipment is designed from these circuit diagrams. In logic tests of VLSI's, path sensitizing methods are widely used to generate test specifications. But the method generates test specifications at a single time only, and can not be directly applied to sequential control software. The basic idea of the proposed method is as follows. Specifications of each logic operator in the diagrams are defined in the software design process. Therefore, test specifications of each operator in the control software can be determined from these specifications, and validity of software can be judged by inspecting all of the operators in the logic circuit diagrams. Candidates for sensitized paths, on which test data for each operator propagates, can be generated by the path sensitizing method. To confirm feasibility of the method, it was experimentally applied to control software in digital control equipment. The program could generate test specifications exactly, and feasibility of the method was confirmed. (orig.) (3 refs., 7 figs.)

  14. Communicative Language Testing: Implications for Computer Based Language Testing in French for Specific Purposes

    Science.gov (United States)

    García Laborda, Jesús; López Santiago, Mercedes; Otero de Juan, Nuria; Álvarez Álvarez, Alfredo

    2014-01-01

    Current evolutions of language testing have led to integrating computers in FSP assessments both in oral and written communicative tasks. This paper deals with two main issues: learners' expectations about the types of questions in FSP computer based assessments and the relation with their own experience. This paper describes the experience of 23…

  15. Modelling of Aquitaine II pipe whipping test with EUROPLEXUS fast dynamics code

    International Nuclear Information System (INIS)

    Potapov, S.

    2003-01-01

    To validate the modelling of multi-physics phenomena with EUROPLEXUS code we considered a pipe whipping problem occurring in thermal hydraulic conditions of a Loss of Coolant Accident in PWR primary circuit. Two numerical fluid-structure interaction (FSI) models, a simplified 'pipe-like' model and a mixed 1D/3D model, were used to simulate both the conditioning phase and a phase of whipping. The results of calculations were compared with existing experimental data. Analysis of numerical results shows that both models give a good prediction of global behaviour of the coupled fluid-structure system, namely for pipe displacements and stresses in the pipe walls, as well as for pressure and velocity in the fluid. By comparison with experimental data, we show that only the mixed EUROPLEXUS model, where the pipe elbow is discretized with shells, allows us to estimate correctly the time history and maximum value of the contact force between the pipe and the obstacle. The 1D model with reduced kinematics (rigid cross section hypothesis) does not allow the correct detection of contact phenomenon. This study shows that the use of mixed numerical models containing simplified and totally 3D parts duly interconnected allows a very efficient and CPU inexpensive numerical analysis which is able to take into account different global and local physical phenomena. (author)

  16. Insights into inner ear-specific gene regulation: epigenetics and non-coding RNAs in inner ear development and regeneration

    Science.gov (United States)

    Avraham, Karen B.

    2016-01-01

    The vertebrate inner ear houses highly specialized sensory organs, tuned to detect and encode sound, head motion and gravity. Gene expression programs under the control of transcription factors orchestrate the formation and specialization of the non-sensory inner ear labyrinth and its sensory constituents. More recently, epigenetic factors and non-coding RNAs emerged as an additional layer of gene regulation, both in inner ear development and disease. In this review, we provide an overview on how epigenetic modifications and non-coding RNAs, in particular microRNAs (miRNAs), influence gene expression and summarize recent discoveries that highlight their critical role in the proper formation of the inner ear labyrinth and its sensory organs. In contrast to non-mammalian vertebrates, adult mammals lack the ability to regenerate inner ear mechano-sensory hair cells. Finally, we discuss recent insights into how epigenetic factors and miRNAs may facilitate, or in the case of mammals, restrict sensory hair cell regeneration. PMID:27836639

  17. Preliminary engineering specifications for a test demonstration multilayer protective barrier cover system

    International Nuclear Information System (INIS)

    Phillips, S.J.; Gilbert, T.W.; Adams, M.R.

    1985-03-01

    This report presents preliminary engineering specifications for a test protective barrier cover system and support radiohydrology facility to be constructed at the Hanford Protective Barrier Test Facility (PBTF). Construction of this test barrier and related radiohydrology facility is part of a continuing effort to provide construction experience and performance evaluation of alternative barrier designs used for long-term isolation of disposed radioactive waste materials. Design specifications given in this report are tentative, based on interim engineering and computer simulation design efforts. Final definitive design specifications and engineering prints will be produced in FY 1986. 6 refs., 10 figs., 1 tab

  18. Project W-314 specific test and evaluation plan for AZ tank farm upgrades

    International Nuclear Information System (INIS)

    Hays, W.H.

    1998-01-01

    The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made by the addition of the SN-631 transfer line from the AZ-O1A pit to the AZ-02A pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation P1 an (TEP). Testing includes Validations and Verifications (e.g., Commercial Grade Item Dedication activities, etc), Factory Tests and Inspections (FTIs), installation tests and inspections, Construction Tests and Inspections (CTIs), Acceptance Test Procedures (ATPs), Pre-Operational Test Procedures (POTPs), and Operational Test Procedures (OTPs). The STEP will be utilized in conjunction with the TEP for verification and validation

  19. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  20. Testing Requirements to Manage Data Exchange Specifications in Enterprise Integration - A Schema Design Quality Focus.

    Energy Technology Data Exchange (ETDEWEB)

    Kulvatunyou, Boonserm [ORNL; Ivezic, Nenad [ORNL; Buhwan, Jeong [POSTECH University, South Korea

    2004-07-01

    In this paper, we describe the requirements to test W3C XML Schema usage when defining message schemas for data exchange in any large and evolving enterprise integration project. We then decompose the XML Schema testing into four (4) aspects including the message schema conformance to the XML Schema specification grammar, the message schema conformance to the XML Schema specification semantics, the message schema conformance to design quality testing, and canonical semantics testing of the message schema. We describe these four testing aspects in some detail and point to other related efforts. We further focus to provide some technical details for the message schema design quality testing. As a future work, we describe the requirements for canonical semantics testing and potential solution approaches. Finally, we describe an implementation architecture for the message schema design quality testing.

  1. Non-coding changes cause sex-specific wing size differences between closely related species of Nasonia

    NARCIS (Netherlands)

    Loehlin, David W.; Oliveira, Deodoro C. S. G.; Edwards, Rachel; Giebel, Jonathan D.; Clark, Michael E.; Cattani, M. Victoria; van de Zande, Louis; Verhulst, Eveline C.; Beukeboom, Leo W.; Munoz-Torres, Monica; Werren, John H.

    The genetic basis of morphological differences among species is still poorly understood. We investigated the genetic basis of sex-specific differences in wing size between two closely related species of Nasonia by positional cloning a major male-specific locus, wing-size1 (ws1). Male wing size

  2. Project W-314 specific test and evaluation plan 241-AN-B valve pit

    International Nuclear Information System (INIS)

    Hays, W.H.

    1998-01-01

    The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AN-B Valve Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation Plan (TEP)

  3. Project W-314 specific test and evaluation plan for 241-AN-A valve pit

    International Nuclear Information System (INIS)

    Hays, W.H.

    1998-01-01

    The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AN-A Valve Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation Plan (TEP)

  4. 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.

  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. 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)

  7. Mining Software Repositories to Study Co-Evolution of Production & Test Code

    NARCIS (Netherlands)

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

    2008-01-01

    Preprint of paper published in: ICST 2008 - Proceedings of the International Conference on Software Testing, Verification, and Validation, 2008; doi:10.1109/ICST.2008.47 Engineering software systems is a multidisciplinary activity, whereby a number of artifacts must be created — and maintained —

  8. 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.

  9. Evaluation of RELAP5 MOD 3.1.1 code with GIRAFFE Test Facility: Phase 1, Step 2 nitrogen venting tests

    International Nuclear Information System (INIS)

    Boyer, B.D.; Slovik, G.C.; Rohatgl, U.S.

    1995-01-01

    The Simplified Boiling Water Reactor (SBWR) proposed by General Electric (GE) is an advanced light water reactor (ALWR) design that utilizes passive safety systems. The PCCS is a series of heat exchangers submerged in water and open to the containment. Since the containment is inerted with nitrogen during normal operation, the PCCS must condense the steam in the presence of noncondensable gases during an accident. To model the transient behavior of the SBWR with a system code, the code should properly simulate the expected phenomena. To validate the applicability of RELAP5 MOD 3.1.1, the data from three Phase 1, Step 2 nitrogen venting tests at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal facility and RELAP5 calculations of these tests were compared. The comparison of the GIRAFFE data against the results from the RELAP5 calculations showed that it can predict condensation and gas purging phenomena occurring in the long-term decay heat rejection phase. In this phase of the transient, condensation in the PCCS is the only means to reject heat from the SBWR containment. In the two tests where the nitrogen purge vent line was at its deepest submergence in the Suppression Pool (SIP), the RELAP5 results mirrored the behavior of the containment pressures and of the water levels in the Horizontal Vent (HV) and the nitrogen purge line tube of the GIRAFFE data. However, in the test with the shallowest purge line submergence, there was appreciable direct contact condensation on the pool surface of the HV despite modeling efforts to deter these phenomena. This surface condensation, unobserved in the GIRAFFE tests, was a major cause of RELAP5 predicting early containment depressurization and the subsequent early rise in HV and nitrogen purge line water levels. The present RELAP5 MOD3.1.1 interfacial heat and mass transfer model does not properly degrade direct contact steam condensation in the presence of noncondensable gases sitting on a pool

  10. Pre-analysis of Phenix End-of-Life Thermal-hydraulic tests with the MARS-LMR Code

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Ha, Kwi Seok; Kwon, Young Min; Chang, Won Pyo; Suk, Su Dong; Lee, Yong Bum

    2009-01-01

    A prototype SFR, PHENIX has been operated by the French Commissariat a l'energie atomique (CEA) and the Electricite de France (EdF) since 1973. Through the successful operation for 35 years, PHENIX has achieved its original objective to demonstrate a fast breeder reactor technology and also played an important role as an irradiation facility for innovative fuels and materials. Since its first operation, PHENIX has accumulated about 4,300 equivalent full power days (EFPDs) of operational experience and it reached its final shutdown in 2009. Before the decommissioning of PHENIX, the CEA started a PHENIX end-of-life (EOL) test program and opened it for international collaboration to share the valuable information from the test. The KAERI joined this program to utilize the unique opportunity to validate its SFR system analysis code, MARS-LMR which will be a basic tool in future SFR development

  11. Revision to dedicated short range communication roadside equipment specification - RSU 4.1.Bench Test Plan.

    Science.gov (United States)

    2017-04-28

    The document describes the overall process for evaluating Dedicated Short Range Communication (DSRC) Roadside Units (RSU) against USDOT RSU Specification 4.1 in preparation for field evaluation. The Test Cases contained in this document only evaluate...

  12. Youth's perceptions of HIV infection risk: a sex-specific test of two ...

    African Journals Online (AJOL)

    Youth's perceptions of HIV infection risk: a sex-specific test of two risk models. ... The analysis is based on data from the 2003 Demographic and Health survey ... multiple partners, Nigeria, risk perception, sexual behaviour, vulnerability to HIV ...

  13. Validity of multiple stress creep recovery test for LADOTD asphalt binder specification.

    Science.gov (United States)

    2010-09-01

    The objectives of this research are to characterize the elastic response of various binders used by LADOTD to determine the feasibility of the Multiple Stress Creep Recovery (MSCR) test to be included in the LADOTD asphalt binder specification and to...

  14. A closure test for time-specific capture-recapture data

    Science.gov (United States)

    Stanley, T.R.; Burnham, K.P.

    1999-01-01

    The assumption of demographic closure in the analysis of capture-recapture data under closed-population models is of fundamental importance. Yet, little progress has been made in the development of omnibus tests of the closure assumption. We present a closure test for time-specific data that, in principle, tests the null hypothesis of closed-population model M(t) against the open-population Jolly-Seber model as a specific alternative. This test is chi-square, and can be decomposed into informative components that can be interpreted to determine the nature of closure violations. The test is most sensitive to permanent emigration and least sensitive to temporary emigration, and is of intermediate sensitivity to permanent or temporary immigration. This test is a versatile tool for testing the assumption of demographic closure in the analysis of capture-recapture data.

  15. Development of blow down and sodium-water reaction jet analysis codes-Validation by sodium-water reaction tests (SWAT-1R)

    International Nuclear Information System (INIS)

    Hiroshi Seino; Akikazu Kurihara; Isao Ono; Koji Jitsu

    2005-01-01

    Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed in order to improve the evaluation method on sodium-water reaction event in the steam generator (SG) of a sodium cooled fast breeder reactor (FBR). The validation analyses by these two codes were carried out using the data of Sodium-Water Reaction Test (SWAT-1R). The following main results have been obtained through this validation: (1) The calculational results by LEAP-BLOW such as internal pressure and water flow rate show good agreement with the results of the SWAT- 1R test. (2) The LEAP-JET code can qualitatively simulate the behavior of sodium-water reaction. However, it is found that the code has tendency to overestimate the maximum temperature of the reaction jet. (authors)

  16. Pre-test calculations for FAL-19 and FAL-20 using the ITHACA code

    International Nuclear Information System (INIS)

    Bradley, S.J.; Ketchell, N.

    1992-08-01

    Falcon is a small scale experimental apparatus, designed to simulate the transport of fission products through the primary circuit and containment of a nuclear power reactor under severe accident conditions. Information gained from the experiments in Falcon will be used to guide and assist in understanding the much larger Phebus-FP experiments. This report presents the results of pre-test calculations performed using ITHACA for the two tests: FAL-19 and FAL-20. Initial calculations were concerned solely with the thermal-hydraulic conditions in the containment while later ones briefly investigated the effect of the injection of an insoluble aerosol into the containment with the same thermal-hydraulic conditions. (author)

  17. Pre-test analysis of ATLAS SBO with RCP seal leakage scenario using MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Quang Huy; Lee, Sang Young; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    This study presents a pre-test calculation for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. Initially, turbine-driven auxfeed water pumps are used. Then, outside cooling water injection method is used for long term cooling. The analysis results would be useful for conducting the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection in future. The pre-test calculation for ATLAS extended SBO with RCP seal leakage and outside cooling water injection scenario is performed. After Fukushima nuclear accident, the capability of coping with the extended station blackout (SBO) becomes important. Many NPPs are applying FLEX approach as main coping strategies for extended SBO scenarios. In FLEX strategies, outside cooling water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. It is worthwhile to examine the soundness of outside cooling water injection method for extended SBO mitigation by both calculation and experimental demonstration. From the calculation results, outside cooling water injection into RCS and SGs is verified as an effective method during extended SBO when RCS and SGs depressurization is sufficiently performed.

  18. 40 CFR 53.51 - Demonstration of compliance with design specifications and manufacturing and test requirements.

    Science.gov (United States)

    2010-07-01

    ... Methods and Class I and Class II Equivalent Methods for PM2.5 or PM10â2.5 § 53.51 Demonstration of... standard specification 8625F, Type II, Class I (reference 4 in appendix A of this subpart) in the same way... specifications and manufacturing and test requirements. 53.51 Section 53.51 Protection of Environment...

  19. Experimental test of host specificity in a behaviour-modifying trematode

    DEFF Research Database (Denmark)

    Hernandez, R.N.; Fredensborg, Brian Lund

    2015-01-01

    Host behavioural modification by parasites is a common and well-documented phenomenon. However, knowledge on the complexity and specificity of the underlying mechanisms is limited, and host specificity among manipulating parasites has rarely been experimentally verified. We tested the hypothesis...

  20. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

  1. The test-retest reliability and criterion validity of a high-intensity, netball-specific circuit test: The Net-Test.

    Science.gov (United States)

    Mungovan, Sean F; Peralta, Paula J; Gass, Gregory C; Scanlan, Aaron T

    2018-04-12

    To examine the test-retest reliability and criterion validity of a high-intensity, netball-specific fitness test. Repeated measures, within-subject design. Eighteen female netball players competing in an international competition completed a trial of the Net-Test, which consists of 14 timed netball-specific movements. Players also completed a series of netball-relevant criterion fitness tests. Ten players completed an additional Net-Test trial one week later to assess test-retest reliability using intraclass correlation coefficient (ICC), typical error of measurement (TEM), and coefficient of variation (CV). The typical error of estimate expressed as CV and Pearson correlations were calculated between each criterion test and Net-Test performance to assess criterion validity. Five movements during the Net-Test displayed moderate ICC (0.84-0.90) and two movements displayed high ICC (0.91-0.93). Seven movements and heart rate taken during the Net-Test held low CV (Test possessed low CV and significant (pTest possesses acceptable reliability for the assessment of netball fitness. Further, the high criterion validity for the Net-Test suggests a range of important netball-specific fitness elements are assessed in combination. Copyright © 2018 Sports Medicine Australia. Published by Elsevier Ltd. All rights reserved.

  2. Enhancing co-operation between AVN, IRSN and GRS: the junior staff pilot project on the comparative testing of IPA codes

    International Nuclear Information System (INIS)

    Hoyos, A. de; Keesmann, S.; Smidts, O.

    2006-01-01

    - Objectives: The project takes place within the framework of the Junior Staff Program of AVN, GRS and IRSN which aims at creating a junior staff network among European TSOs. The objective of this project is to apply integrated performance assessment (IPA) tools used by AVN, IRSN and GRS to two generic and simplified models (Bure site in France and Mol site in Belgium) for disposal systems in argillaceous formations. The comparison of the results from different codes applied to the disposal systems of the two mentioned sites aims at a better understanding of the confinement capabilities of the considered geological formations and of the IPA methodology in general. The incentive is a common understanding of approaches developed by each partner and the improvement of this expertise. More specifically, this pilot project aims at enhancing exchanges of views and mutual experiences in the field of understanding major safety functions. - Tools and Methods: A new code for the assessment of barrier systems in argillaceous formations has only recently been developed at GRS, as in the past such formations played a minor role as a possible hosting environment for a repository in Germany. The project also serves as a test case for this code. The considered disposal systems are defined on the basis of the concepts and data available for Mol and Bure. The program packages used for the performance assessment calculations are: HYDRUS-1D with source term module (AVN), GoldSim (IRSN) and EMOS-modules CLAYPOS and CHET (GRS). While the coupling of HYDRUS-1D with a source term module and the EMOS-modules are FORTRAN77- coded programs specifically developed for the simulation of parts of a barrier system of a final repository, GoldSim is a general purpose simulation environment with an integrated graphical user interface for modelling and data output. Models realized in GoldSim are flexible and can be easily adapted to new requirements. The software also offers an intrinsic

  3. Development of an improved species specific PCR test for detection of Haemophilus parasuis

    DEFF Research Database (Denmark)

    Angen, Øystein; Oliveira, Simone; Ahrens, Peter

    2007-01-01

    , the present PCR test was found to be 100% species specific for H. parasuis, in contrast to the PCR test of Oliveira et al., which also tested positive for strains belonging to A. indolicus, A. porcinus, and A. minor, species commonly occurring in the upper respiratory tract. However, when the PCR test...... with representatives of H. parasuis. The test was further evaluated on 55 clinical samples from 16 Danish pigs suspected for being infected with H. parasuis, showing polyserositis or septicemia at autopsy as well as on 492 nasal swabs. The test was compared with the performance of a PCR test earlier published...... by Oliveira et al. [Oliveira, S., Galina, L., Pijoan, C., 2001. Development of a PCR test to diagnose Haemophilus parasuis infections. J. Vet. Diagn. Invest. 13, 495-501]. The sensitivity of the present PCR test was found to be slightly lower when applied on clinical samples from diseased pigs and 10-fold...

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

    International Nuclear Information System (INIS)

    Tanaka, Mitsugu

    1978-01-01

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

  5. German boiler and pressure vessel codes and standards: materials, manufacture, testing, equipment, erection and operation

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

    The methods by which the safety objectives on the operation of steam boilers and pressure vessels in Germany can be reached are set out in Technical Rules which are compiled and established in technical committees. Typical applications are described in the Technical Rules. A chart shows how the laws, provisions and Technical Rules for the sections 'steam boiler plant' and 'pressure vessels' are interlinked. This chapter concentrates on legal aspects, materials, manufacture, testing, erection and operation of boilers and pressure vessels in Germany. (U.K.)

  6. Melt/concrete interactions: the Sandia experimental program, model development, and code comparison test

    International Nuclear Information System (INIS)

    Powers, D.A.; Muir, J.F.

    1979-01-01

    High temperature melt/concrete interactions have been studied both experimentally and analytically at Sandia under sponsorship of Reactor Safety Research of the US Nuclear Regulatory Commission. The purpose of these studies has been to develop an understanding of these interactions suitable for risk assessment. Results of the experimental program are summarized and a computer model of melt/concrete interactions is described. A melt/concrete interaction test that will allow this and other models of the interaction to be compared is also described

  7. Dynamic benchmarking of simulation codes

    International Nuclear Information System (INIS)

    Henry, R.E.; Paik, C.Y.; Hauser, G.M.

    1996-01-01

    output includes a plot of the MAAP calculation and the plant data. For the large integral experiments, a major part, but not all of the MAAP code is needed. These use an experiment specific benchmark routine that includes all of the information and boundary conditions for performing the calculation, as well as the information of which parts of MAAP are unnecessary and can be 'bypassed'. Lastly, the separate effects tests only require a few MAAP routines. These are exercised through their own specific benchmark routine that includes the experiment specific information and boundary conditions. This benchmark routine calls the appropriate MAAP routines from the source code, performs the calculations, including integration where necessary and provide the comparison between the MAAP calculation and the experimental observations. (author)

  8. Vertical jumping tests in volleyball: reliability, validity, and playing-position specifics.

    Science.gov (United States)

    Sattler, Tine; Sekulic, Damir; Hadzic, Vedran; Uljevic, Ognjen; Dervisevic, Edvin

    2012-06-01

    Vertical jumping is known to be important in volleyball, and jumping performance tests are frequently studied for their reliability and validity. However, most studies concerning jumping in volleyball have dealt with standard rather than sport-specific jumping procedures and tests. The aims of this study, therefore, were (a) to determine the reliability and factorial validity of 2 volleyball-specific jumping tests, the block jump (BJ) test and the attack jump (AJ) test, relative to 2 frequently used and systematically validated jumping tests, the countermovement jump test and the squat jump test and (b) to establish volleyball position-specific differences in the jumping tests and simple anthropometric indices (body height [BH], body weight, and body mass index [BMI]). The BJ was performed from a defensive volleyball position, with the hands positioned in front of the chest. During an AJ, the players used a 2- to 3-step approach and performed a drop jump with an arm swing followed by a quick vertical jump. A total of 95 high-level volleyball players (all men) participated in this study. The reliability of the jumping tests ranged from 0.97 to 0.99 for Cronbach's alpha coefficients, from 0.93 to 0.97 for interitem correlation coefficients and from 2.1 to 2.8 for coefficients of variation. The highest reliability was found for the specific jumping tests. The factor analysis extracted one significant component, and all of the tests were highly intercorrelated. The analysis of variance with post hoc analysis showed significant differences between 5 playing positions in some of the jumping tests. In general, receivers had a greater jumping capacity, followed by libero players. The differences in jumping capacities should be emphasized vis-a-vis differences in the anthropometric measures of players, where middle hitters had higher BH and body weight, followed by opposite hitters and receivers, with no differences in the BMI between positions.

  9. Sensitivity and specificity of the 3-item memory test in the assessment of post traumatic amnesia.

    Science.gov (United States)

    Andriessen, Teuntje M J C; de Jong, Ben; Jacobs, Bram; van der Werf, Sieberen P; Vos, Pieter E

    2009-04-01

    To investigate how the type of stimulus (pictures or words) and the method of reproduction (free recall or recognition after a short or a long delay) affect the sensitivity and specificity of a 3-item memory test in the assessment of post traumatic amnesia (PTA). Daily testing was performed in 64 consecutively admitted traumatic brain injured patients, 22 orthopedically injured patients and 26 healthy controls until criteria for resolution of PTA were reached. Subjects were randomly assigned to a test with visual or verbal stimuli. Short delay reproduction was tested after an interval of 3-5 minutes, long delay reproduction was tested after 24 hours. Sensitivity and specificity were calculated over the first 4 test days. The 3-word test showed higher sensitivity than the 3-picture test, while specificity of the two tests was equally high. Free recall was a more effortful task than recognition for both patients and controls. In patients, a longer delay between registration and recall resulted in a significant decrease in the number of items reproduced. Presence of PTA is best assessed with a memory test that incorporates the free recall of words after a long delay.

  10. Wind-US Code Physical Modeling Improvements to Complement Hypersonic Testing and Evaluation

    Science.gov (United States)

    Georgiadis, Nicholas J.; Yoder, Dennis A.; Towne, Charles S.; Engblom, William A.; Bhagwandin, Vishal A.; Power, Greg D.; Lankford, Dennis W.; Nelson, Christopher C.

    2009-01-01

    This report gives an overview of physical modeling enhancements to the Wind-US flow solver which were made to improve the capabilities for simulation of hypersonic flows and the reliability of computations to complement hypersonic testing. The improvements include advanced turbulence models, a bypass transition model, a conjugate (or closely coupled to vehicle structure) conduction-convection heat transfer capability, and an upgraded high-speed combustion solver. A Mach 5 shock-wave boundary layer interaction problem is used to investigate the benefits of k- s and k-w based explicit algebraic stress turbulence models relative to linear two-equation models. The bypass transition model is validated using data from experiments for incompressible boundary layers and a Mach 7.9 cone flow. The conjugate heat transfer method is validated for a test case involving reacting H2-O2 rocket exhaust over cooled calorimeter panels. A dual-mode scramjet configuration is investigated using both a simplified 1-step kinetics mechanism and an 8-step mechanism. Additionally, variations in the turbulent Prandtl and Schmidt numbers are considered for this scramjet configuration.

  11. Screening for Specific Language Impairment in Preschool Children: Evaluating a Screening Procedure Including the Token Test

    Science.gov (United States)

    Willinger, Ulrike; Schmoeger, Michaela; Deckert, Matthias; Eisenwort, Brigitte; Loader, Benjamin; Hofmair, Annemarie; Auff, Eduard

    2017-01-01

    Specific language impairment (SLI) comprises impairments in receptive and/or expressive language. Aim of this study was to evaluate a screening for SLI. 61 children with SLI (SLI-children, age-range 4-6 years) and 61 matched typically developing controls were tested for receptive language ability (Token Test-TT) and for intelligence (Wechsler…

  12. Latent tuberculosis in HIV positive, diagnosed by the M. tuberculosis specific interferon-gamma test

    DEFF Research Database (Denmark)

    Brock, Inger; Ruhwald, Morten; Lundgren, Bettina

    2006-01-01

    Although tuberculosis (TB) is a minor problem in Denmark, severe and complicated cases occur in HIV positive. Since the new M. tuberculosis specific test for latent TB, the QuantiFERON-TB In-Tube test (QFT-IT) became available the patients in our clinic have been screened for the presence of latent...

  13. Latent tuberculosis in HIV positive, diagnosed by the M. tuberculosis specific interferon-gamma test

    DEFF Research Database (Denmark)

    Brock, Inger; Ruhwald, Morten; Lundgren, Bettina

    2006-01-01

    BACKGROUND: Although tuberculosis (TB) is a minor problem in Denmark, severe and complicated cases occur in HIV positive. Since the new M. tuberculosis specific test for latent TB, the QuantiFERON-TB In-Tube test (QFT-IT) became available the patients in our clinic have been screened...

  14. Test Specification of A1-2 Test for OECD-ATLAS Project

    International Nuclear Information System (INIS)

    Ryu, Sung Uk; Kim, Seok; Euh, Dong-Jin

    2014-01-01

    According to Sateesh et al., the model for boiling on non-horizontal surfaces should consider microlayer evaporation and transient conduction owing to the sliding of bubbles, as shown in Eq. (1) q tot = (q me +q tc )x st +(q mes +q tcs )x s + q nc , (1)where q tot is the total heat flux, q me and q tc are the microlayer evaporation and transient conduction heat flux from a stationary bubble, q mes and q tcs are the microlayer evaporation and transient conduction heat flux owing to the sliding bubbles, q nc is the natural convection heat flux, x st and x s are constants determined by the area ratio parameter R defined as the ratio of area available per nucleation site to the projected area of the bubble at departure. In a model of wall heat flux partitioning, the microlayer evaporation from sliding bubbles q mes can be defined by four sub-models, i.e., the bubble departure diameter d d , bubble lift-off diameter d 1 , bubble departure frequency f, and active nucleation site density n b , as shown in Eq. (2) q mes =1/6, (2) where is density of the vapour, and h fg is the specific latent heat. Among these sub-models, this paper focuses on the bubble lift-off diameter. Situ et al. stated that the bubble lift-off diameter, which is the bubble size when a bubble detaches from the heater surface, can be different from the bubble departure size, which is the bubble size when a bubble detaches from the nucleation site. There have been a number of works performed on the departure and lift-off diameters of the bubbles generated on non-horizontal surfaces: Schomann, Luke and Gonfleo, Luke (study on the horizontal tube) Cornwell and Schuller, Situ et al., and Cho et al. (study on the vertical surface). Although there are many useful models to predict the departure and lift-off diameters of the bubbles generated on non-horizontal surfaces, the previous researchers did not deal with the bubble lift-off diameter model applicable on a horizontal tube. The boiling phenomena on the

  15. The one-dimensional transport code CHET2, taking into account nonlinear, element-specific equilibrium sorption

    International Nuclear Information System (INIS)

    Luehrmann, L.; Noseck, U.

    1996-03-01

    While the verification report on CHET1 primarily focused on aspects such as the correctness of algorithms with respect to the modeling of advection, dispersion and diffusion, the report in hand is intended to primarily deal with nonlinear sorption and numerical sorption modeling. Another aspect discussed is the correct treatment of decay within established radioactive decay chains. First, the physical fundamentals are explained of the processes determining the radionuclide transport in the cap rock, and hence are the basis of the program discussed. The numeric algorithms the CHET2 code is based are explained, showing the details of realisation and the function of the various defaults and corrections. The iterative coupling of transport and sorption computation is illustrated by means of a program flowchart. Furthermore, the actvities for verification of the program are explained, as well as qualitative effects of computations assuming concentration-dependent sorption. The computation of the decay within decay chains is verified, and application programming using nonlinear sorption isotherms as well as the entire process of transport calculations with CHET2 are shown. (orig./DG) [de

  16. Development of a Culture Specific Critical Thinking Ability Test and Using It as a Supportive Diagnostic Test for Giftedness

    Science.gov (United States)

    Köksal, Mustafa Serdar

    2016-01-01

    The purposes of this study were to develop a culture specific critical thinking ability test for 6, 7, and 8. grade students in Turkey and to use it as an assessment instrument for giftedness. For these purposes, item pool involving 22 items was formed by writing items focusing on the current and common events presented in (Turkish) media from…

  17. Validation of a field test for the non-invasive determination of badminton specific aerobic performance

    Science.gov (United States)

    Wonisch, M; Hofmann, P; Schwaberger, G; von Duvillard, S P; Klein, W

    2003-01-01

    Aim: To develop a badminton specific test to determine on court aerobic and anaerobic performance. Method: The test was evaluated by using a lactate steady state test. Seventeen male competitive badminton players (mean (SD) age 26 (8) years, weight 74 (10) kg, height 179 (7) cm) performed an incremental field test on the badminton court to assess the heart rate turn point (HRTP) and the individual physical working capacity (PWCi) at 90% of measured maximal heart rate (HRmax). All subjects performed a 20 minute steady state test at a workload just below the PWCi. Results: Significant correlations (pbadminton is possible without HRTP determination. PMID:12663351

  18. Specificity tests of an oligonucleotide probe against food-outbreak salmonella for biosensor detection

    Science.gov (United States)

    Chen, I.-H.; Horikawa, S.; Xi, J.; Wikle, H. C.; Barbaree, J. M.; Chin, B. A.

    2017-05-01

    Phage based magneto-elastic (ME) biosensors have been shown to be able to rapidly detect Salmonella in various food systems to serve food pathogen monitoring purposes. In this ME biosensor platform, the free-standing strip-shaped magneto-elastic sensor is the transducer and the phage probe that recognizes Salmonella in food serves as the bio-recognition element. According to Sorokulova et al. at 2005, a developed oligonucleotide probe E2 was reported to have high specificity to Salmonella enterica Typhimurium. In the report, the specificity tests were focused in most of Enterobacterace groups outside of Salmonella family. Here, to understand the specificity of phage E2 to different Salmonella enterica serotypes within Salmonella Family, we further tested the specificity of the phage probe to thirty-two Salmonella serotypes that were present in the major foodborne outbreaks during the past ten years (according to Centers for Disease Control and Prevention). The tests were conducted through an Enzyme linked Immunosorbent Assay (ELISA) format. This assay can mimic probe immobilized conditions on the magnetoelastic biosensor platform and also enable to study the binding specificity of oligonucleotide probes toward different Salmonella while avoiding phage/ sensor lot variations. Test results confirmed that this oligonucleotide probe E2 was high specific to Salmonella Typhimurium cells but showed cross reactivity to Salmonella Tennessee and four other serotypes among the thirty-two tested Salmonella serotypes.

  19. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    International Nuclear Information System (INIS)

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

  20. Improved E-ELT subsystem and component specifications, thanks to M1 test facility

    Science.gov (United States)

    Dimmler, M.; Marrero, J.; Leveque, S.; Barriga, Pablo; Sedghi, B.; Kornweibel, N.

    2014-07-01

    During the last 2 years ESO has operated the "M1 Test Facility", a test stand consisting of a representative section of the E-ELT primary mirror equipped with 4 complete prototype segment subunits including sensors, actuators and control system. The purpose of the test facility is twofold: it serves to study and get familiar with component and system aspects like calibration, alignment and handling procedures and suitable control strategies on real hardware long before the primary mirror (hereafter M1) components are commissioned. Secondly, and of major benefit to the project, it offered the possibility to evaluate component and subsystem performance and interface issues in a system context in such detail, that issues could be identified early enough to feed back into the subsystem and component specifications. This considerably reduces risk and cost of the production units and allows refocusing the project team on important issues for the follow-up of the production contracts. Experiences are presented in which areas the results of the M1 Test Facility particularly helped to improve subsystem specifications and areas, where additional tests were adopted independent of the main test facility. Presented are the key experiences of the M1 Test Facility which lead to improved specifications or identified the need for additional testing outside of the M1 Test Facility.

  1. CodonTest: modeling amino acid substitution preferences in coding sequences.

    Directory of Open Access Journals (Sweden)

    Wayne Delport

    2010-08-01

    Full Text Available Codon models of evolution have facilitated the interpretation of selective forces operating on genomes. These models, however, assume a single rate of non-synonymous substitution irrespective of the nature of amino acids being exchanged. Recent developments have shown that models which allow for amino acid pairs to have independent rates of substitution offer improved fit over single rate models. However, these approaches have been limited by the necessity for large alignments in their estimation. An alternative approach is to assume that substitution rates between amino acid pairs can be subdivided into rate classes, dependent on the information content of the alignment. However, given the combinatorially large number of such models, an efficient model search strategy is needed. Here we develop a Genetic Algorithm (GA method for the estimation of such models. A GA is used to assign amino acid substitution pairs to a series of rate classes, where is estimated from the alignment. Other parameters of the phylogenetic Markov model, including substitution rates, character frequencies and branch lengths are estimated using standard maximum likelihood optimization procedures. We apply the GA to empirical alignments and show improved model fit over existing models of codon evolution. Our results suggest that current models are poor approximations of protein evolution and thus gene and organism specific multi-rate models that incorporate amino acid substitution biases are preferred. We further anticipate that the clustering of amino acid substitution rates into classes will be biologically informative, such that genes with similar functions exhibit similar clustering, and hence this clustering will be useful for the evolutionary fingerprinting of genes.

  2. Using MathWorks' Simulink® and Real-Time Workshop® Code Generator to Produce Attitude Control Test and Flight Code

    OpenAIRE

    Salada, Mark; Dellinger, Wayne

    1998-01-01

    This paper describes the use of a commercial product, MathWorks' RealTime Workshop® (RTW), to generate actual flight code for NASA's Thermosphere, Ionosphere, Mesosphere Energetics and Dynamics (TIMED) mission. The Johns Hopkins University Applied Physics Laboratory is handling the design and construction of this satellite for NASA. As TIMED is scheduled to launch in May of the year 2000, software development for both ground and flight systems are well on their way. However, based on experien...

  3. Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

    1978-07-01

    The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

  4. 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.

  5. The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

  6. Evaluation of sensitivity and specificity of bone marrow trephine biopsy tests in an Indian teaching hospital

    Directory of Open Access Journals (Sweden)

    Sima Chauhan

    2018-06-01

    Full Text Available Introduction: Bone marrow aspiration (BMA and bone marrow biopsy (BMB is an indispensable diagnostic tool for evaluating haematological and non-haematological disorders and patient follow-up in present era. We have compared the advantages of trephine biopsy over bone marrow aspiration in these patients. Aim and objective: To evaluate sensitivity and specificity of trephine biopsy test for haematological and non haematological disorder patients in comparison to bone marrow aspiration test. Materials and method: In this 1 year prospective study (June 2014–May 2015, we evaluated the haematological and non-haematological disorder patients by BMA and BMB (aided with I.H.C. when ever needed. The sensitivity and specificity of the tests were calculated. Results: Among, final 504 hemotological/non haematological disorder patients, 416 cases were diagnosed (+ve in BMA test, where as it was 494 in BMB test and with chi2 test it was highly significant as p = 0.0001. It was concluded that True positive cases were 416, True negative were 9 cases, false negative 78 cases and false positive was in one case only. The sensitivity and specificity of bone marrow trephine biopsy test was 84% and 90% respectively. Conclusion: BMB (aided with I.H.C is a gold standard test for detecting different haematological and non hamatological disorders. In our study the sensitivity and specificity of BMB test was 84% and 90% respectively. When performed in association with BMA in the same sitting, significantly augments the chances of reaching a correct diagnosis. Keywords: Bone marrow trephine biopsy, Bone marrow aspiration, Sensitivity, Specificity

  7. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  8. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  9. Histone modification profiles are predictive for tissue/cell-type specific expression of both protein-coding and microRNA genes

    Directory of Open Access Journals (Sweden)

    Zhang Michael Q

    2011-05-01

    Full Text Available Abstract Background Gene expression is regulated at both the DNA sequence level and through modification of chromatin. However, the effect of chromatin on tissue/cell-type specific gene regulation (TCSR is largely unknown. In this paper, we present a method to elucidate the relationship between histone modification/variation (HMV and TCSR. Results A classifier for differentiating CD4+ T cell-specific genes from housekeeping genes using HMV data was built. We found HMV in both promoter and gene body regions to be predictive of genes which are targets of TCSR. For example, the histone modification types H3K4me3 and H3K27ac were identified as the most predictive for CpG-related promoters, whereas H3K4me3 and H3K79me3 were the most predictive for nonCpG-related promoters. However, genes targeted by TCSR can be predicted using other type of HMVs as well. Such redundancy implies that multiple type of underlying regulatory elements, such as enhancers or intragenic alternative promoters, which can regulate gene expression in a tissue/cell-type specific fashion, may be marked by the HMVs. Finally, we show that the predictive power of HMV for TCSR is not limited to protein-coding genes in CD4+ T cells, as we successfully predicted TCSR targeted genes in muscle cells, as well as microRNA genes with expression specific to CD4+ T cells, by the same classifier which was trained on HMV data of protein-coding genes in CD4+ T cells. Conclusion We have begun to understand the HMV patterns that guide gene expression in both tissue/cell-type specific and ubiquitous manner.

  10. Sport-specific fitness testing and intervention for an adolescent with cerebral palsy: a case report.

    Science.gov (United States)

    Kenyon, Lisa K; Sleeper, Mark D; Tovin, Melissa M

    2010-01-01

    This case report describes the development, implementation, and outcomes of a fitness-related intervention program that addressed the sport-specific goals of an adolescent with cerebral palsy. The participant in this case was a 16-year-old African American male with spastic diplegia. The participant joined his high school wrestling team and asked to focus his physical therapy on interventions that would improve his wrestling performance. An examination was performed using the muscle power sprint test, the 10 x 5-m sprint test, strength tests, the 10-m shuttle run test, and the Gross Motor Function Measure. The intervention consisted of interval training, which focused on the demands of wrestling. Scores on all tests and measures were higher after the intervention. The outcomes of this case report seem to support the use of a fitness-related intervention program for addressing the sport-specific goals of an adolescent with cerebral palsy.

  11. Evaluation of a draft standard on performance specifications for health physics instrumentation: results for environmental tests

    International Nuclear Information System (INIS)

    Kenoyer, J.L.; Swinth, K.L.; Mashburn, K.R.; Selby, J.M.

    1984-06-01

    Draft ANSI Standard N42.17 on performance specifications for health physics instrumentation is currently being evaluated by the Pacific Northwest Laboratory. Evaluation is performed by testing a cross-section of currently available instruments with testing procedures based on specifications of the standard and then determining the degree of conformance to the various elements of the proposed standard. Data will be presented on the performance of a cross-section of beta-gamma survey instruments under various environmental tests. Test results that will be presented include temperature effects, humidity effects, radio frequency (r.f.) susceptibility, ambient pressure effects, vibration effects, and shock effects. Tests performed to date show that most instruments will meet the temperature, humidity, and ambient pressure tests. A large variability is noted among instruments from the same or different vendors. Preliminary r.f. susceptibility tests have shown large artificial responses at some frequencies for specific instruments. The presentation will also include a discussion of procedures used in the testing and weaknesses identified in the proposed standard

  12. Comparison of liver fibrosis blood tests developed for HCV with new specific tests in HIV/HCV co-infection.

    Science.gov (United States)

    Calès, Paul; Halfon, Philippe; Batisse, Dominique; Carrat, Fabrice; Perré, Philippe; Penaranda, Guillaume; Guyader, Dominique; d'Alteroche, Louis; Fouchard-Hubert, Isabelle; Michelet, Christian; Veillon, Pascal; Lambert, Jérôme; Weiss, Laurence; Salmon, Dominique; Cacoub, Patrice

    2010-08-01

    We compared 5 non-specific and 2 specific blood tests for liver fibrosis in HCV/HIV co-infection. Four hundred and sixty-seven patients were included into derivation (n=183) or validation (n=284) populations. Within these populations, the diagnostic target, significant fibrosis (Metavir F > or = 2), was found in 66% and 72% of the patients, respectively. Two new fibrosis tests, FibroMeter HICV and HICV test, were constructed in the derivation population. Unadjusted AUROCs in the derivation population were: APRI: 0.716, Fib-4: 0.722, Fibrotest: 0.778, Hepascore: 0.779, FibroMeter: 0.783, HICV test: 0.822, FibroMeter HICV: 0.828. AUROCs adjusted on classification and distribution of fibrosis stages in a reference population showed similar values in both populations. FibroMeter, FibroMeter HICV and HICV test had the highest correct classification rates in F0/1 and F3/4 (which account for high predictive values): 77-79% vs. 70-72% in the other tests (p=0.002). Reliable individual diagnosis based on predictive values > or = 90% distinguished three test categories: poorly reliable: Fib-4 (2.4% of patients), APRI (8.9%); moderately reliable: Fibrotest (25.4%), FibroMeter (26.6%), Hepascore (30.2%); acceptably reliable: HICV test (40.2%), FibroMeter HICV (45.6%) (ptests). FibroMeter HICV classified all patients into four reliable diagnosis intervals ( or =F1, > or =F2) with an overall accuracy of 93% vs. 79% (pfibrosis. Tests designed for HCV infections are less effective in HIV/HCV infections. A specific test, like FibroMeter HICV, was the most interesting test for diagnostic accuracy, correct classification profile, and a reliable diagnosis. With reliable diagnosis intervals, liver biopsy can therefore be avoided in all patients. Copyright 2010 European Association for the Study of the Liver. Published by Elsevier B.V. All rights reserved.

  13. Implementation of the International Code of Practice on Dosimetry in Diagnostic Radiology (TRS 457): Review of Test Results

    International Nuclear Information System (INIS)

    2011-01-01

    In 2007, the IAEA published Dosimetry in Diagnostic Radiology: An International Code of Practice (IAEA Technical Reports Series No. 457). This publication recommends procedures for calibration and dosimetric measurement for the attainment of standardized dosimetry. It also addresses requirements both in standards dosimetry laboratories, especially Secondary Standards Dosimetry Laboratories (SSDLs), and in clinical centres for radiology, as found in most hospitals. The implementation of TRS No. 457 decreases the uncertainty in the dosimetry of diagnostic radiology beams and provides Member States with a unified and consistent framework for dosimetry in diagnostic radiology, which previously did not exist. A coordinated research project (CRP E2.10.06) was established in order to provide practical guidance to professionals at SSDLs and to clinical medical physicists on the implementation of TRS No. 457. This includes the calibration of radiological dosimetry instrumentation, the dissemination of calibration coefficients to clinical centres and the establishment of dosimetric measurement processes in clinical settings. The main goals of the CRP were to: Test the procedures recommended in TRS No. 457 for calibration of radiation detectors in different types of diagnostic beams and measuring instruments for varying diagnostic X ray modalities; Test the clinical dosimetry procedures, including the use of phantoms and patient dose surveys; Report on the practical implementation of TRS No. 457 at both SSDLs and hospital sites. Testing of TRS No. 457 was performed by a group of medical physicists from hospitals and SSDLs from various institutions worldwide

  14. 200 Area Treated Effluent Disposal Facility operational test specification. Revision 2

    International Nuclear Information System (INIS)

    Crane, A.F.

    1995-01-01

    This document identifies the test specification and test requirements for the 200 Area Treated Effluent Disposal Facility (200 Area TEDF) operational testing activities. These operational testing activities, when completed, demonstrate the functional, operational and design requirements of the 200 Area TEDF have been met. The technical requirements for operational testing of the 200 Area TEDF are defined by the test requirements presented in Appendix A. These test requirements demonstrate the following: pump station No.1 and associated support equipment operate both automatically and manually; pump station No. 2 and associated support equipment operate both automatically and manually; water is transported through the collection and transfer lines to the disposal ponds with no detectable leakage; the disposal ponds accept flow from the transfer lines with all support equipment operating as designed; and the control systems operate and status the 200 Area TEDF including monitoring of appropriate generator discharge parameters

  15. Ca(2+) coding and decoding strategies for the specification of neural and renal precursor cells during development.

    Science.gov (United States)

    Moreau, Marc; Néant, Isabelle; Webb, Sarah E; Miller, Andrew L; Riou, Jean-François; Leclerc, Catherine

    2016-03-01

    During embryogenesis, a rise in intracellular Ca(2+) is known to be a widespread trigger for directing stem cells towards a specific tissue fate, but the precise Ca(2+) signalling mechanisms involved in achieving these pleiotropic effects are still poorly understood. In this review, we compare the Ca(2+) signalling events that appear to be one of the first steps in initiating and regulating both neural determination (neural induction) and kidney development (nephrogenesis). We have highlighted the necessary and sufficient role played by Ca(2+) influx and by Ca(2+) transients in the determination and differentiation of pools of neural or renal precursors. We have identified new Ca(2+) target genes involved in neural induction and we showed that the same Ca(2+) early target genes studied are not restricted to neural tissue but are also present in other tissues, principally in the pronephros. In this review, we also described a mechanism whereby the transcriptional control of gene expression during neurogenesis and nephrogenesis might be directly controlled by Ca(2+) signalling. This mechanism involves members of the Kcnip family such that a change in their binding properties to specific DNA sites is a result of Ca(2+) binding to EF-hand motifs. The different functions of Ca(2+) signalling during these two events illustrate the versatility of Ca(2+) as a second messenger. Copyright © 2015 Elsevier Ltd. All rights reserved.

  16. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  17. Circular codes revisited: a statistical approach.

    Science.gov (United States)

    Gonzalez, D L; Giannerini, S; Rosa, R

    2011-04-21

    In 1996 Arquès and Michel [1996. A complementary circular code in the protein coding genes. J. Theor. Biol. 182, 45-58] discovered the existence of a common circular code in eukaryote and prokaryote genomes. Since then, circular code theory has provoked great interest and underwent a rapid development. In this paper we discuss some theoretical issues related to the synchronization properties of coding sequences and circular codes with particular emphasis on the problem of retrieval and maintenance of the reading frame. Motivated by the theoretical discussion, we adopt a rigorous statistical approach in order to try to answer different questions. First, we investigate the covering capability of the whole class of 216 self-complementary, C(3) maximal codes with respect to a large set of coding sequences. The results indicate that, on average, the code proposed by Arquès and Michel has the best covering capability but, still, there exists a great variability among sequences. Second, we focus on such code and explore the role played by the proportion of the bases by means of a hierarchy of permutation tests. The results show the existence of a sort of optimization mechanism such that coding sequences are tailored as to maximize or minimize the coverage of circular codes on specific reading frames. Such optimization clearly relates the function of circular codes with reading frame synchronization. Copyright © 2011 Elsevier Ltd. All rights reserved.

  18. 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

  19. Analysis of Phenix End-of-Life asymmetry test with multi-dimensional pool modeling of MARS-LMR code

    International Nuclear Information System (INIS)

    Jeong, H.-Y.; Ha, K.-S.; Choi, C.-W.; Park, M.-G.

    2015-01-01

    Highlights: • Pool behaviors under asymmetrical condition in an SFR were evaluated with MARS-LMR. • The Phenix asymmetry test was analyzed one-dimensionally and multi-dimensionally. • One-dimensional modeling has limitation to predict the cold pool temperature. • Multi-dimensional modeling shows improved prediction of stratification and mixing. - Abstract: The understanding of complicated pool behaviors and its modeling is essential for the design and safety analysis of a pool-type Sodium-cooled Fast Reactor. One of the remarkable recent efforts on the study of pool thermal–hydraulic behaviors is the asymmetrical test performed as a part of Phenix End-of-Life tests by the CEA. To evaluate the performance of MARS-LMR code, which is a key system analysis tool for the design of an SFR in Korea, in the prediction of thermal hydraulic behaviors during an asymmetrical condition, the Phenix asymmetry test is analyzed with MARS-LMR in the present study. Pool regions are modeled with two different approaches, one-dimensional modeling and multi-dimensional one, and the prediction results are analyzed to identify the appropriateness of each modeling method. The prediction with one-dimensional pool modeling shows a large deviation from the measured data at the early stage of the test, which suggests limitations to describe the complicated thermal–hydraulic phenomena. When the pool regions are modeled multi-dimensionally, the prediction gives improved results quite a bit. This improvement is explained by the enhanced modeling of pool mixing with the multi-dimensional modeling. On the basis of the results from the present study, it is concluded that an accurate modeling of pool thermal–hydraulics is a prerequisite for the evaluation of design performance and safety margin quantification in the future SFR developments

  20. Performance of an automatic dose control system for CT. Specifications and basic phantom tests

    Energy Technology Data Exchange (ETDEWEB)

    Nagel, H.D. [Wissenschaft und Technik fuer die Radiolgoe, Dr. HD Nagel, Buchholz (Germany); Stumpp, P.; Kahn, T.; Gosch, D. [Universitaetsklinikum Leipzig (Germany). Klinik und Poliklinik fuer Diagnostische und Interventionelle Radiologie

    2011-01-15

    Purpose: To assess the performance and to provide more detailed insight into the characteristics and limitations of devices for automatic dose control (ADC) in CT. Materials and Methods: A comprehensive study on DoseRight 2.0, the ADC system provided by Philips for its Brilliance CT scanners, was conducted. Phantom tests were carried out on a 64-slice scanner (Brilliance 64) using assorted quality control (QC) phantoms that allowed verification of the basic specifications. If feasible, the findings were verified by model calculations based on known specifications. Results: For all tests, the dose reductions and modulation characteristics fully met the values expected from the specifications. Adverse effects due to increased image noise were only moderate as a result of the 'adequate noise system' design that employs comparatively gentle modulation, and the additional use of adaptive filtration. Conclusion: Simple tests with QC phantoms allow evaluation of the most relevant characteristics of devices for ADC in CT. (orig.)

  1. Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

    International Nuclear Information System (INIS)

    Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

    2011-01-01

    Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

  2. Pre-screening Discussions and Prostate-Specific Antigen Testing for Prostate Cancer Screening.

    Science.gov (United States)

    Li, Jun; Zhao, Guixiang; Hall, Ingrid J

    2015-08-01

    For many men, the net benefit of prostate cancer screening with prostate-specific antigen (PSA) tests may be small. Many major medical organizations have issued recommendations for prostate cancer screening, stressing the need for shared decision making before ordering a test. The purpose of this study is to better understand associations between discussions about benefits and harms of PSA testing and uptake of the test among men aged ≥40 years. Associations between pre-screening discussions and PSA testing were examined using self-reported data from the 2012 Behavioral Risk Factor Surveillance System. Unadjusted prevalence of PSA testing was estimated and AORs were calculated using logistic regression in 2014. The multivariate analysis showed that men who had ever discussed advantages of PSA testing only or discussed both advantages and disadvantages were more likely, respectively, to report having had a test within the past year than men who had no discussions (ptesting with their healthcare providers were more likely (AOR=2.75, 95% CI=2.00, 3.79) to report getting tested than men who had no discussions. Discussions of the benefits or harms of PSA testing are positively associated with increased uptake of the test. Given the conflicting recommendations for prostate cancer screening and increasing importance of shared decision making, this study points to the need for understanding how pre-screening discussions are being conducted in clinical practice and the role played by patients' values and preferences in decisions about PSA testing. Published by Elsevier Inc.

  3. Sensitivity and specificity of parallel or serial serological testing for detection of canine Leishmania infection

    Directory of Open Access Journals (Sweden)

    Mauro Maciel de Arruda

    2016-01-01

    Full Text Available In Brazil, human and canine visceral leishmaniasis (CVL caused byLeishmania infantum has undergone urbanisation since 1980, constituting a public health problem, and serological tests are tools of choice for identifying infected dogs. Until recently, the Brazilian zoonoses control program recommended enzyme-linked immunosorbent assays (ELISA and indirect immunofluorescence assays (IFA as the screening and confirmatory methods, respectively, for the detection of canine infection. The purpose of this study was to estimate the accuracy of ELISA and IFA in parallel or serial combinations. The reference standard comprised the results of direct visualisation of parasites in histological sections, immunohistochemical test, or isolation of the parasite in culture. Samples from 98 cases and 1,327 noncases were included. Individually, both tests presented sensitivity of 91.8% and 90.8%, and specificity of 83.4 and 53.4%, for the ELISA and IFA, respectively. When tests were used in parallel combination, sensitivity attained 99.2%, while specificity dropped to 44.8%. When used in serial combination (ELISA followed by IFA, decreased sensitivity (83.3% and increased specificity (92.5% were observed. Serial testing approach improved specificity with moderate loss in sensitivity. This strategy could partially fulfill the needs of public health and dog owners for a more accurate diagnosis of CVL.

  4. Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

    2012-07-01

    The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

  5. Improvement of western blot test specificity for detecting equine serum antibodies to Sarcocystis neurona.

    Science.gov (United States)

    Rossano, M G; Mansfield, L S; Kaneene, J B; Murphy, A J; Brown, C M; Schott, H C; Fox, J C

    2000-01-01

    Equine protozoal myeloencephalitis (EPM) is a neurological disease of horses and ponies caused by the apicomplexan protozoan parasite Sarcocystis neurona. The purposes of this study were to develop the most stringent criteria possible for a positive test result, to estimate the sensitivity and specificity of the EPM Western blot antibody test, and to assess the ability of bovine antibodies to Sarcocystis cruzi to act as a blocking agent to minimize false-positive results in the western blot test for S. neurona. Sarcocystis neurona merozoites harvested from equine dermal cell culture were heat denatured, and the proteins were separated by sodium dodecyl sulfate-polyacrylamide gel electrophoresis in a 12-20% linear gradient gel. Separated proteins were electrophoretically transferred to polyvinylidene fluoride membranes and blocked in 1% bovine serum albumin and 0.5% Tween-Tris-buffered saline. Serum samples from 6 horses with S. neurona infections (confirmed by culture from neural tissue) and 57 horses without infections (horses from the Eastern Hemisphere, where S. neurona does not exist) were tested by Western blot. Horses from both groups had reactivity to the 62-, 30-, 16-, 13-, 11-, 10.5-, and 10-kD bands. Testing was repeated with another step. Blots were treated with bovine S. cruzi antibodies prior to loading the equine samples. After this modification of the Western blot test, positive infection status was significantly associated with reactivity to the 30- and 16-kD bands (Pblot had a sample sensitivity of 100% and sample specificity of 98%. It is concluded that the specificity of the Western blot test is improved by blocking proteins not specific to S. neurona and using reactivity to the 30- and 16-kD bands as the criterion for a positive test.

  6. Development of Integral Effect Test Facility P and ID and Technical Specification for SMART Fluid System

    International Nuclear Information System (INIS)

    Lee, Sang Il; Jung, Y. H.; Yang, H. J.; Song, S. Y.; Han, O. J.; Lee, B. J.; Kim, Y. A.; Lim, J. H.; Park, K. W.; Kim, N. G.

    2010-01-01

    SMART integral test loop is the thermal hydraulic test facility with a high pressure and temperature for simulating the major systems of the prototype reactor, SMART-330. The objective of this project is to conduct the basic design for constructing SMART ITL. The major results of this project include a series of design documents, technical specifications and P and ID. The results can be used as the fundamental materials for making the detailed design which is essential for manufacturing and installing SMART ITL

  7. The level 1 and 2 specification for parallel benchmark and a benchmark test of scalar-parallel computer SP2 based on the specifications

    International Nuclear Information System (INIS)

    Orii, Shigeo

    1998-06-01

    A benchmark specification for performance evaluation of parallel computers for numerical analysis is proposed. Level 1 benchmark, which is a conventional type benchmark using processing time, measures performance of computers running a code. Level 2 benchmark proposed in this report is to give the reason of the performance. As an example, scalar-parallel computer SP2 is evaluated with this benchmark specification in case of a molecular dynamics code. As a result, the main causes to suppress the parallel performance are maximum band width and start-up time of communication between nodes. Especially the start-up time is proportional not only to the number of processors but also to the number of particles. (author)

  8. Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2010-05-15

    Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)

  9. Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2010-06-15

    Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)

  10. Assessment of Specificity of the Badcamp Agility Test for Badminton Players

    Directory of Open Access Journals (Sweden)

    de França Bahia Loureiro Luiz

    2017-06-01

    Full Text Available The Badcamp agility test was created to evaluate agility of badminton players. The Badcamp is a valid and reliable test, however, a doubt about the need for the use of this test exists as simpler tests could provide similar information about agility in badminton players. Thus, the aim of this study was to examine the specificity of the Badcamp, comparing the performance of badminton players and athletes from other sports in the Badcamp and the shuttle run agility test (SRAT. Sixty-four young male and female athletes aged between 14 and 16 years participated in the study. They were divided into 4 groups of 16 according to their sport practices: badminton, tennis, team sport (basketball and volleyball, and track and field. We compared the groups in both tests, the Badcamp and SRAT. The results revealed that the group of badminton players was faster compared to all other groups in the Badcamp. However, in the SRAT there were no differences among groups composed of athletes from open skill sports (e.g., badminton, tennis, and team sports, and a considerable reduction of the difference between badminton players and track and field athletes. Thus, we concluded that the Badcamp test is a specific agility test for badminton players and should be considered in evaluating athletes of this sport modality.

  11. Assessment of Specificity of the Badcamp Agility test for Badminton Players.

    Science.gov (United States)

    de França Bahia Loureiro, Luiz; Costa Dias, Mário Oliveira; Cremasco, Felipe Couto; da Silva, Maicon Guimarães; de Freitas, Paulo Barbosa

    2017-06-01

    The Badcamp agility test was created to evaluate agility of badminton players. The Badcamp is a valid and reliable test, however, a doubt about the need for the use of this test exists as simpler tests could provide similar information about agility in badminton players. Thus, the aim of this study was to examine the specificity of the Badcamp, comparing the performance of badminton players and athletes from other sports in the Badcamp and the shuttle run agility test (SRAT). Sixty-four young male and female athletes aged between 14 and 16 years participated in the study. They were divided into 4 groups of 16 according to their sport practices: badminton, tennis, team sport (basketball and volleyball), and track and field. We compared the groups in both tests, the Badcamp and SRAT. The results revealed that the group of badminton players was faster compared to all other groups in the Badcamp. However, in the SRAT there were no differences among groups composed of athletes from open skill sports (e.g., badminton, tennis, and team sports), and a considerable reduction of the difference between badminton players and track and field athletes. Thus, we concluded that the Badcamp test is a specific agility test for badminton players and should be considered in evaluating athletes of this sport modality.

  12. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  13. A study of longwave radiation codes for climate studies: Validation with ARM observations and tests in general circulation models

    International Nuclear Information System (INIS)

    Ellingson, R.G.; Baer, F.

    1993-01-01

    This report summarizes the activities of our group to meet our stated objectives. The report is divided into sections entitled: Radiation Model Testing Activities, General Circulation Model Testing Activities, Science Team Activities, and Publications, Presentations and Meetings. The section on Science Team Activities summarizes our participation with the science team to further advance the observation and modeling programs. Appendix A lists graduate students supported, and post-doctoral appointments during the project. Reports on the activities during each of the first two years are included as Appendix B. Significant progress has been made in: determining the ability of line-by-line radiation models to calculate the downward longwave flux at the surface; determining the uncertainties in calculated the downwelling radiance and flux at the surface associated with the use of different proposed profiling techniques; intercomparing clear-sky radiance and flux observations with calculations from radiation codes from different climate models; determining the uncertainties associated with estimating N* from surface longwave flux observations; and determining the sensitivity of model calculations to different formulations of the effects of finite sized clouds

  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. Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

    Science.gov (United States)

    Sabotinov, Luben; Chevrier, Patrick

    The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

  16. Relationship between the Handball-Specific Complex-Test and Intermittent Field Test performance in professional players.

    Science.gov (United States)

    Hermassi, Souhail; Hoffmeyer, Birgit; Irlenbusch, Lars; Fieseler, Georg; Noack, Frank; Delank, Karl-Stefan; Gabbett, Tim J; Souhaiel Chelly, Mohamed; Schwesig, René

    2018-01-01

    We investigated the relationship between the Handball Complex-Test (HBCT) and two selected field performance tests (the repeated sprint ability [RSA], and the Yo-Yo Intermittent Recovery Test) in elite handball players. Nineteen handball players (age: 25.7±5.1 years) were drawn from the First Professional German League. The HBCT consists of four activity series (AS): agility parcours, defensive action, sprint (10 m, 20 m) and throw-on-goal parcours; these activities were completed twice, with five active pauses of 30-35 s, and a follow-up of recovery over the subsequent 10 minutes. The RSA comprised 6 x (15+15 m) sprints starting every 20 s; scoring noted best time (RSAbest), total time (RSATT) and decrement (RSAdec). In the Yo-Yo Intermittent Recover, we recorded the total distance covered (TD). Heart rates (HR) were recorded throughout and recovery was assessed for measurements immediately post-test (R0) and 10 minutes after completing the test (R10). A strong correlation was found between HBCT and fastest 10 m and 20 m RSA sprint times (r=0.811, r=0.815, respectively). Also, the HBCT total 10 m and 20 m sprint times showed a strong positive association with RSATT (r=0.70; r=0.63, respectively), and the RSA heart rate post-test was strongly correlated with the HBCT heart rate after round two (r=0.865). Data from the match-specific HBCT Test shows a strong positive association with other more generic intermittent field test measurements. These observations support the validity of using the generic tests to monitor current fitness and responses to training in team handball players.

  17. Colon specific CODES based Piroxicam tablet for colon targeting: statistical optimization, in vivo roentgenography and stability assessment.

    Science.gov (United States)

    Singh, Amit Kumar; Pathak, Kamla

    2015-03-01

    This study was aimed to statistically optimize CODES™ based Piroxicam (PXM) tablet for colon targeting. A 3(2) full factorial design was used for preparation of core tablet that was subsequently coated to get CODES™ based tablet. The experimental design of core tablets comprised of two independent variables: amount of lactulose and PEG 6000, each at three different levels and the dependent variable was %CDR at 12 h. The core tablets were evaluated for pharmacopoeial and non-pharmacopoeial test and coated with optimized levels of Eudragit E100 followed by HPMC K15 and finally with Eudragit S100. The in vitro drug release study of F1-F9 was carried out by change over media method (0.1 N HCl buffer, pH 1.2, phosphate buffer, pH 7.4 and phosphate buffer, pH 6.8 with enzyme β-galactosidase 120 IU) to select optimized formulation F9 that was subjected to in vivo roentgenography. Roentgenography study corroborated the in vitro performance, thus providing the proof of concept. The experimental design was validated by extra check point formulation and Diffuse Reflectance Spectroscopy revealed absence of any interaction between drug and formulation excipients. The shelf life of F9 was deduced as 12 months. Conclusively, colon targeted CODES™ technology based PXM tablets were successfully optimized and its potential of colon targeting was validated by roentgenography.

  18. Sensitivity, Specificity, and Positivity Predictors of the Pneumococcal Urinary Antigen Test in Community-Acquired Pneumonia.

    Science.gov (United States)

    Molinos, Luis; Zalacain, Rafael; Menéndez, Rosario; Reyes, Soledad; Capelastegui, Alberto; Cillóniz, Catia; Rajas, Olga; Borderías, Luis; Martín-Villasclaras, Juan J; Bello, Salvador; Alfageme, Inmaculada; Rodríguez de Castro, Felipe; Rello, Jordi; Ruiz-Manzano, Juan; Gabarrús, Albert; Musher, Daniel M; Torres, Antoni

    2015-10-01

    Detection of the C-polysaccharide of Streptococcus pneumoniae in urine by an immune-chromatographic test is increasingly used to evaluate patients with community-acquired pneumonia. We assessed the sensitivity and specificity of this test in the largest series of cases to date and used logistic regression models to determine predictors of positivity in patients hospitalized with community-acquired pneumonia. We performed a multicenter, prospective, observational study of 4,374 patients hospitalized with community-acquired pneumonia. The urinary antigen test was done in 3,874 cases. Pneumococcal infection was diagnosed in 916 cases (21%); 653 (71%) of these cases were diagnosed exclusively by the urinary antigen test. Sensitivity and specificity were 60 and 99.7%, respectively. Predictors of urinary antigen positivity were female sex; heart rate≥125 bpm, systolic blood pressureantibiotic treatment; pleuritic chest pain; chills; pleural effusion; and blood urea nitrogen≥30 mg/dl. With at least six of all these predictors present, the probability of positivity was 52%. With only one factor present, the probability was only 12%. The urinary antigen test is a method with good sensitivity and excellent specificity in diagnosing pneumococcal pneumonia, and its use greatly increased the recognition of community-acquired pneumonia due to S. pneumoniae. With a specificity of 99.7%, this test could be used to direct simplified antibiotic therapy, thereby avoiding excess costs and risk for bacterial resistance that result from broad-spectrum antibiotics. We also identified predictors of positivity that could increase suspicion for pneumococcal infection or avoid the unnecessary use of this test.

  19. Comparison of 2015 Medicare relative value units for gender-specific procedures: Gynecologic and gynecologic-oncologic versus urologic CPT coding. Has time healed gender-worth?

    Science.gov (United States)

    Benoit, M F; Ma, J F; Upperman, B A

    2017-02-01

    In 1992, Congress implemented a relative value unit (RVU) payment system to set reimbursement for all procedures covered by Medicare. In 1997, data supported that a significant gender bias existed in reimbursement for gynecologic compared to urologic procedures. The present study was performed to compare work and total RVU's for gender specific procedures effective January 2015 and to evaluate if time has healed the gender-based RVU worth. Using the 2015 CPT codes, we compared work and total RVU's for 50 pairs of gender specific procedures. We also evaluated 2015 procedure related provider compensation. The groups were matched so that the procedures were anatomically similar. We also compared 2015 to 1997 RVU and fee schedules. Evaluation of work RVU's for the paired procedures revealed that in 36 cases (72%), male vs female procedures had a higher wRVU and tRVU. For total fee/reimbursement, 42 (84%) male based procedures were compensated at a higher rate than the paired female procedures. On average, male specific surgeries were reimbursed at an amount that was 27.67% higher for male procedures than for female-specific surgeries. Female procedure based work RVU's have increased minimally from 1997 to 2015. Time and effort have trended towards resolution of some gender-related procedure worth discrepancies but there are still significant RVU and compensation differences that should be further reviewed and modified as surgical time and effort highly correlate. Copyright © 2016. Published by Elsevier Inc.

  20. 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.