WorldWideScience

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. 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)

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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

  14. 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

  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. 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

  17. 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.

  18. 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

  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. 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. 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...

  2. 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

  3. 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.

  4. 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.

  5. 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.

  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 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.

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. 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.

  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. 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

  16. 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.

  17. 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)

  18. Automated Testing Infrastructure and Result Comparison for Geodynamics Codes

    Science.gov (United States)

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

    2013-12-01

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

  19. 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. 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

  1. 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.

  2. 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)

  3. 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.)

  4. 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

  5. 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.

  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. 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

  8. 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

  9. 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.

  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. 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

  12. 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)

  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. 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

  15. 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

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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)

  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. 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)

  8. 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

  9. 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

  10. 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]....

  11. 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)

  12. 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 ...

  13. 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.)

  14. 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

  15. 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)

  16. On-line monitoring and inservice inspection in codes

    International Nuclear Information System (INIS)

    Bartonicek, J.; Zaiss, W.; Bath, H.R.

    1999-01-01

    The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

  17. 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…

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. 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...

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. Visualizing code and coverage changes for code review

    NARCIS (Netherlands)

    Oosterwaal, Sebastiaan; van Deursen, A.; De Souza Coelho, R.; Sawant, A.A.; Bacchelli, A.

    2016-01-01

    One of the tasks of reviewers is to verify that code modifications are well tested. However, current tools offer little support in understanding precisely how changes to the code relate to changes to the tests. In particular, it is hard to see whether (modified) test code covers the changed code.

  14. SPACE Code Assessment for FLECHT Test

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  15. 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

  16. Coding in pigeons: Multiple-coding versus single-code/default strategies.

    Science.gov (United States)

    Pinto, Carlos; Machado, Armando

    2015-05-01

    To investigate the coding strategies that pigeons may use in a temporal discrimination tasks, pigeons were trained on a matching-to-sample procedure with three sample durations (2s, 6s and 18s) and two comparisons (red and green hues). One comparison was correct following 2-s samples and the other was correct following both 6-s and 18-s samples. Tests were then run to contrast the predictions of two hypotheses concerning the pigeons' coding strategies, the multiple-coding and the single-code/default. According to the multiple-coding hypothesis, three response rules are acquired, one for each sample. According to the single-code/default hypothesis, only two response rules are acquired, one for the 2-s sample and a "default" rule for any other duration. In retention interval tests, pigeons preferred the "default" key, a result predicted by the single-code/default hypothesis. In no-sample tests, pigeons preferred the key associated with the 2-s sample, a result predicted by multiple-coding. Finally, in generalization tests, when the sample duration equaled 3.5s, the geometric mean of 2s and 6s, pigeons preferred the key associated with the 6-s and 18-s samples, a result predicted by the single-code/default hypothesis. The pattern of results suggests the need for models that take into account multiple sources of stimulus control. © Society for the Experimental Analysis of Behavior.

  17. 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.

  18. 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.

  19. 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

  20. 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 ...

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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…

  18. 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.

  19. 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)

  20. 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.

  1. 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)

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. Administrative database code accuracy did not vary notably with changes in disease prevalence.

    Science.gov (United States)

    van Walraven, Carl; English, Shane; Austin, Peter C

    2016-11-01

    Previous mathematical analyses of diagnostic tests based on the categorization of a continuous measure have found that test sensitivity and specificity varies significantly by disease prevalence. This study determined if the accuracy of diagnostic codes varied by disease prevalence. We used data from two previous studies in which the true status of renal disease and primary subarachnoid hemorrhage, respectively, had been determined. In multiple stratified random samples from the two previous studies having varying disease prevalence, we measured the accuracy of diagnostic codes for each disease using sensitivity, specificity, and positive and negative predictive value. Diagnostic code sensitivity and specificity did not change notably within clinically sensible disease prevalence. In contrast, positive and negative predictive values changed significantly with disease prevalence. Disease prevalence had no important influence on the sensitivity and specificity of diagnostic codes in administrative databases. Copyright © 2016 Elsevier Inc. All rights reserved.

  8. 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

  9. 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.

  10. 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'

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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

  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. 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

  19. 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)

  20. Lesson learned from the application to LOBI tests of CATHARE and RELAP5 codes

    International Nuclear Information System (INIS)

    Ambrosini, W.; D'Auria, F.; Galassi, G.M.

    1992-01-01

    The Dipt. di Costruzioni Meccaniche e Nucleari has participated to the LOBI project since its very beginning, contributing to almost all the international activities in this field, such as task group meetings, International Standards Problems, Seminars, etc. System codes like RELAP4/MOD6, RELAP5/MOD1, RELAP5/MOD1-EUR, RELAP5/MOD2, CATHARE 1 and CATHARE 2 were applied to the design and post test evaluation of a wide series of both LOBI/MOD1 and LOBI/MOD2 experiments, including Large Break LOCAs, Small and Intermediate Break LOCAs, long lasting transients and characterization tests. The LOBI data base demonstrated its usefulness in assessing capabilities and limitations of these codes and in qualifying a code use strategy. (author)

  1. 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)

  2. 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

  3. 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 ...

  4. 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

  5. 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)

  6. 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

  7. 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…

  8. 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.

  9. 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.

  10. OM Code Requirements For MOVs -- OMN-1 and Appendix III

    Energy Technology Data Exchange (ETDEWEB)

    Kevin G. DeWall

    2011-08-01

    The purpose or scope of the ASME OM Code is to establish the requirements for pre-service and in-service testing of nuclear power plant components to assess their operational readiness. For MOVs this includes those that perform a specific function in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, and mitigating the consequences of an accident. This paper will present a brief history of industry and regulatory activities related to MOVs and the development of Code requirements to address weaknesses in earlier versions of the OM Code. The paper will discuss the MOV requirements contained in the 2009 version of ASME OM Code, specifically Mandatory Appendix III and OMN-1, Revision 1.

  11. OM Code Requirements For MOVs -- OMN-1 and Appendix III

    International Nuclear Information System (INIS)

    DeWall, Kevin G.

    2011-01-01

    The purpose or scope of the ASME OM Code is to establish the requirements for pre-service and in-service testing of nuclear power plant components to assess their operational readiness. For MOVs this includes those that perform a specific function in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, and mitigating the consequences of an accident. This paper will present a brief history of industry and regulatory activities related to MOVs and the development of Code requirements to address weaknesses in earlier versions of the OM Code. The paper will discuss the MOV requirements contained in the 2009 version of ASME OM Code, specifically Mandatory Appendix III and OMN-1, Revision 1.

  12. 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.

  13. Automatic code generation in practice

    DEFF Research Database (Denmark)

    Adam, Marian Sorin; Kuhrmann, Marco; Schultz, Ulrik Pagh

    2016-01-01

    -specific language to specify those requirements and to allow for generating a safety-enforcing layer of code, which is deployed to the robot. The paper at hand reports experiences in practically applying code generation to mobile robots. For two cases, we discuss how we addressed challenges, e.g., regarding weaving......Mobile robots often use a distributed architecture in which software components are deployed to heterogeneous hardware modules. Ensuring the consistency with the designed architecture is a complex task, notably if functional safety requirements have to be fulfilled. We propose to use a domain...... code generation into proprietary development environments and testing of manually written code. We find that a DSL based on the same conceptual model can be used across different kinds of hardware modules, but a significant adaptation effort is required in practical scenarios involving different kinds...

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  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. 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

  20. 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

  1. 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....

  2. 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.

  3. 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.)

  4. 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.

  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. 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

  7. Lawrence Livermore National Laboratory Probabilistic Seismic Hazard Codes Validation

    International Nuclear Information System (INIS)

    Savy, J B

    2003-01-01

    Probabilistic Seismic Hazard Analysis (PSHA) is a methodology that estimates the likelihood that various levels of earthquake-caused ground motion will be exceeded at a given location in a given future time-period. LLNL has been developing the methodology and codes in support of the Nuclear Regulatory Commission (NRC) needs for reviews of site licensing of nuclear power plants, since 1978. A number of existing computer codes have been validated and still can lead to ranges of hazard estimates in some cases. Until now, the seismic hazard community had not agreed on any specific method for evaluation of these codes. The Earthquake Engineering Research Institute (EERI) and the Pacific Engineering Earthquake Research (PEER) center organized an exercise in testing of existing codes with the aim of developing a series of standard tests that future developers could use to evaluate and calibrate their own codes. Seven code developers participated in the exercise, on a voluntary basis. Lawrence Livermore National laboratory participated with some support from the NRC. The final product of the study will include a series of criteria for judging of the validity of the results provided by a computer code. This EERI/PEER project was first planned to be completed by June of 2003. As the group neared completion of the tests, the managing team decided that new tests were necessary. As a result, the present report documents only the work performed to this point. It demonstrates that the computer codes developed by LLNL perform all calculations correctly and as intended. Differences exist between the results of the codes tested, that are attributed to a series of assumptions, on the parameters and models, that the developers had to make. The managing team is planning a new series of tests to help in reaching a consensus on these assumptions

  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. 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.

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. Turbine Internal and Film Cooling Modeling For 3D Navier-Stokes Codes

    Science.gov (United States)

    DeWitt, Kenneth; Garg Vijay; Ameri, Ali

    2005-01-01

    The aim of this research project is to make use of NASA Glenn on-site computational facilities in order to develop, validate and apply aerodynamic, heat transfer, and turbine cooling models for use in advanced 3D Navier-Stokes Computational Fluid Dynamics (CFD) codes such as the Glenn-" code. Specific areas of effort include: Application of the Glenn-HT code to specific configurations made available under Turbine Based Combined Cycle (TBCC), and Ultra Efficient Engine Technology (UEET) projects. Validating the use of a multi-block code for the time accurate computation of the detailed flow and heat transfer of cooled turbine airfoils. The goal of the current research is to improve the predictive ability of the Glenn-HT code. This will enable one to design more efficient turbine components for both aviation and power generation. The models will be tested against specific configurations provided by NASA Glenn.

  1. 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…

  2. 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)

  3. Improvement of MARS code reflood model

    International Nuclear Information System (INIS)

    Hwang, Moonkyu; Chung, Bub-Dong

    2011-01-01

    A specifically designed heat transfer model for the reflood process which normally occurs at low flow and low pressure was originally incorporated in the MARS code. The model is essentially identical to that of the RELAP5/MOD3.3 code. The model, however, is known to have under-estimated the peak cladding temperature (PCT) with earlier turn-over. In this study, the original MARS code reflood model is improved. Based on the extensive sensitivity studies for both hydraulic and wall heat transfer models, it is found that the dispersed flow film boiling (DFFB) wall heat transfer is the most influential process determining the PCT, whereas the interfacial drag model most affects the quenching time through the liquid carryover phenomenon. The model proposed by Bajorek and Young is incorporated for the DFFB wall heat transfer. Both space grid and droplet enhancement models are incorporated. Inverted annular film boiling (IAFB) is modeled by using the original PSI model of the code. The flow transition between the DFFB and IABF, is modeled using the TRACE code interpolation. A gas velocity threshold is also added to limit the top-down quenching effect. Assessment calculations are performed for the original and modified MARS codes for the Flecht-Seaset test and RBHT test. Improvements are observed in terms of the PCT and quenching time predictions in the Flecht-Seaset assessment. In case of the RBHT assessment, the improvement over the original MARS code is found marginal. A space grid effect, however, is clearly seen from the modified version of the MARS code. (author)

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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)

  12. Algorithms and computer codes for atomic and molecular quantum scattering theory

    International Nuclear Information System (INIS)

    Thomas, L.

    1979-01-01

    This workshop has succeeded in bringing up 11 different coupled equation codes on the NRCC computer, testing them against a set of 24 different test problems and making them available to the user community. These codes span a wide variety of methodologies, and factors of up to 300 were observed in the spread of computer times on specific problems. A very effective method was devised for examining the performance of the individual codes in the different regions of the integration range. Many of the strengths and weaknesses of the codes have been identified. Based on these observations, a hybrid code has been developed which is significantly superior to any single code tested. Thus, not only have the original goals been fully met, the workshop has resulted directly in an advancement of the field. All of the computer programs except VIVS are available upon request from the NRCC. Since an improved version of VIVS is contained in the hybrid program, VIVAS, it was not made available for distribution. The individual program LOGD is, however, available. In addition, programs which compute the potential energy matrices of the test problems are also available. The software library names for Tests 1, 2 and 4 are HEH2, LICO, and EN2, respectively

  13. 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.

  14. 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.

  15. Test quality

    International Nuclear Information System (INIS)

    Hartley, R.S.; Keller, A.E.

    1992-01-01

    Inservice testing of safety-related components at nuclear power plants is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section 11 of the Code state test method and frequency requirements for pumps and valves, respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test's ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing

  16. 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.

  17. 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

  18. 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.

  19. Nuclear model codes and related software distributed by the OECD/NEA Data Bank

    International Nuclear Information System (INIS)

    Sartori, E.

    1993-01-01

    Software and data for nuclear energy applications is acquired, tested and distributed by several information centres; in particular, relevant computer codes are distributed internationally by the OECD/NEA Data Bank (France) and by ESTSC and EPIC/RSIC (United States). This activity is coordinated among the centres and is extended outside the OECD area through an arrangement with the IAEA. This article covers more specifically the availability of nuclear model codes and also those codes which further process their results into data sets needed for specific nuclear application projects. (author). 2 figs

  20. 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)

  1. 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...

  2. DLLExternalCode

    Energy Technology Data Exchange (ETDEWEB)

    2014-05-14

    DLLExternalCode is the a general dynamic-link library (DLL) interface for linking GoldSim (www.goldsim.com) with external codes. The overall concept is to use GoldSim as top level modeling software with interfaces to external codes for specific calculations. The DLLExternalCode DLL that performs the linking function is designed to take a list of code inputs from GoldSim, create an input file for the external application, run the external code, and return a list of outputs, read from files created by the external application, back to GoldSim. Instructions for creating the input file, running the external code, and reading the output are contained in an instructions file that is read and interpreted by the DLL.

  3. 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.

  4. 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

  5. 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

  6. Report number codes

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.N. (ed.)

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name.

  7. Report number codes

    International Nuclear Information System (INIS)

    Nelson, R.N.

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name

  8. 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

  9. 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

  10. TPASS: a gamma-ray spectrum analysis and isotope identification computer code

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1981-03-01

    The gamma-ray spectral data-reduction and analysis computer code TPASS is described. This computer code is used to analyze complex Ge(Li) gamma-ray spectra to obtain peak areas corrected for detector efficiencies, from which are determined gamma-ray yields. These yields are compared with an isotope gamma-ray data file to determine the contributions to the observed spectrum from decay of specific radionuclides. A complete FORTRAN listing of the code and a complex test case are given

  11. Software Certification - Coding, Code, and Coders

    Science.gov (United States)

    Havelund, Klaus; Holzmann, Gerard J.

    2011-01-01

    We describe a certification approach for software development that has been adopted at our organization. JPL develops robotic spacecraft for the exploration of the solar system. The flight software that controls these spacecraft is considered to be mission critical. We argue that the goal of a software certification process cannot be the development of "perfect" software, i.e., software that can be formally proven to be correct under all imaginable and unimaginable circumstances. More realistically, the goal is to guarantee a software development process that is conducted by knowledgeable engineers, who follow generally accepted procedures to control known risks, while meeting agreed upon standards of workmanship. We target three specific issues that must be addressed in such a certification procedure: the coding process, the code that is developed, and the skills of the coders. The coding process is driven by standards (e.g., a coding standard) and tools. The code is mechanically checked against the standard with the help of state-of-the-art static source code analyzers. The coders, finally, are certified in on-site training courses that include formal exams.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. Discussion on LDPC Codes and Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the progress that the workgroup on Low-Density Parity-Check (LDPC) for space link coding. The workgroup is tasked with developing and recommending new error correcting codes for near-Earth, Lunar, and deep space applications. Included in the presentation is a summary of the technical progress of the workgroup. Charts that show the LDPC decoder sensitivity to symbol scaling errors are reviewed, as well as a chart showing the performance of several frame synchronizer algorithms compared to that of some good codes and LDPC decoder tests at ESTL. Also reviewed is a study on Coding, Modulation, and Link Protocol (CMLP), and the recommended codes. A design for the Pseudo-Randomizer with LDPC Decoder and CRC is also reviewed. A chart that summarizes the three proposed coding systems is also presented.

  17. 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)

  18. 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.

  19. 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)

  20. 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...

  1. 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.

  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. 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

  4. 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)

  5. 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

  6. Synthesizing Certified Code

    Science.gov (United States)

    Whalen, Michael; Schumann, Johann; Fischer, Bernd

    2002-01-01

    Code certification is a lightweight approach to demonstrate software quality on a formal level. Its basic idea is to require producers to provide formal proofs that their code satisfies certain quality properties. These proofs serve as certificates which can be checked independently. Since code certification uses the same underlying technology as program verification, it also requires many detailed annotations (e.g., loop invariants) to make the proofs possible. However, manually adding theses annotations to the code is time-consuming and error-prone. We address this problem by combining code certification with automatic program synthesis. We propose an approach to generate simultaneously, from a high-level specification, code and all annotations required to certify generated code. Here, we describe a certification extension of AUTOBAYES, a synthesis tool which automatically generates complex data analysis programs from compact specifications. AUTOBAYES contains sufficient high-level domain knowledge to generate detailed annotations. This allows us to use a general-purpose verification condition generator to produce a set of proof obligations in first-order logic. The obligations are then discharged using the automated theorem E-SETHEO. We demonstrate our approach by certifying operator safety for a generated iterative data classification program without manual annotation of the code.

  7. Test quality

    International Nuclear Information System (INIS)

    Hartley, R.S.; Keller, A.E.

    1992-01-01

    This document discusses inservice testing of safety-related components at nuclear power plants which is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section XI of the Code state test method and frequency requirements for pumps and valves respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test's ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing

  8. The fast code

    Energy Technology Data Exchange (ETDEWEB)

    Freeman, L.N.; Wilson, R.E. [Oregon State Univ., Dept. of Mechanical Engineering, Corvallis, OR (United States)

    1996-09-01

    The FAST Code which is capable of determining structural loads on a flexible, teetering, horizontal axis wind turbine is described and comparisons of calculated loads with test data are given at two wind speeds for the ESI-80. The FAST Code models a two-bladed HAWT with degrees of freedom for blade bending, teeter, drive train flexibility, yaw, and windwise and crosswind tower motion. The code allows blade dimensions, stiffnesses, and weights to differ and models tower shadow, wind shear, and turbulence. Additionally, dynamic stall is included as are delta-3 and an underslung rotor. Load comparisons are made with ESI-80 test data in the form of power spectral density, rainflow counting, occurrence histograms, and azimuth averaged bin plots. It is concluded that agreement between the FAST Code and test results is good. (au)

  9. 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

  10. 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...

  11. 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

  12. 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…

  13. 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

  14. 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

  15. 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....

  16. COMPUTATION FORMAT computer codes X4TOC4 and PLOTC4. Implementing and Testing on a Personal Computer

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1987-05-01

    This document describes the contents of the diskette containing the COMPUTATION FORMAT codes X4TOC4 and PLOTC4 by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a single diskette. (author)

  17. 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. RBMK-LOCA-Analyses with the ATHLET-Code

    Energy Technology Data Exchange (ETDEWEB)

    Petry, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Kurfuerstendamm, Berlin (Germany); Domoradov, A.; Finjakin, A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    1995-09-01

    The scientific technical cooperation between Germany and Russia includes the area of adaptation of several German codes for the Russian-designed RBMK-reactor. One point of this cooperation is the adaptation of the Thermal-Hydraulic code ATHLET (Analyses of the Thermal-Hydraulics of LEaks and Transients), for RBMK-specific safety problems. This paper contains a short description of a RBMK-1000 reactor circuit. Furthermore, the main features of the thermal-hydraulic code ATHLET are presented. The main assumptions for the ATHLET-RBMK model are discussed. As an example for the application, the results of test calculations concerning a guillotine type rupture of a distribution group header are presented and discussed, and the general analysis conditions are described. A comparison with corresponding RELAP-calculations is given. This paper gives an overview on some problems posed and experience by application of Western best-estimate codes for RBMK-calculations.

  19. 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)

  20. 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.

  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. Quality Improvement of MARS Code and Establishment of Code Coupling

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Kim, Kyung Doo

    2010-04-01

    The improvement of MARS code quality and coupling with regulatory auditing code have been accomplished for the establishment of self-reliable technology based regulatory auditing system. The unified auditing system code was realized also by implementing the CANDU specific models and correlations. As a part of the quality assurance activities, the various QA reports were published through the code assessments. The code manuals were updated and published a new manual which describe the new models and correlations. The code coupling methods were verified though the exercise of plant application. The education-training seminar and technology transfer were performed for the code users. The developed MARS-KS is utilized as reliable auditing tool for the resolving the safety issue and other regulatory calculations. The code can be utilized as a base technology for GEN IV reactor applications

  3. 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)

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

    International Nuclear Information System (INIS)

    Reznik, L.

    1994-01-01

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

  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

    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

  6. 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.

  7. High-Fidelity Coding with Correlated Neurons

    Science.gov (United States)

    da Silveira, Rava Azeredo; Berry, Michael J.

    2014-01-01

    Positive correlations in the activity of neurons are widely observed in the brain. Previous studies have shown these correlations to be detrimental to the fidelity of population codes, or at best marginally favorable compared to independent codes. Here, we show that positive correlations can enhance coding performance by astronomical factors. Specifically, the probability of discrimination error can be suppressed by many orders of magnitude. Likewise, the number of stimuli encoded—the capacity—can be enhanced more than tenfold. These effects do not necessitate unrealistic correlation values, and can occur for populations with a few tens of neurons. We further show that both effects benefit from heterogeneity commonly seen in population activity. Error suppression and capacity enhancement rest upon a pattern of correlation. Tuning of one or several effective parameters can yield a limit of perfect coding: the corresponding pattern of positive correlation leads to a ‘lock-in’ of response probabilities that eliminates variability in the subspace relevant for stimulus discrimination. We discuss the nature of this pattern and we suggest experimental tests to identify it. PMID:25412463

  8. 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)

  9. Science Library of Test Items. Volume Twelve. Mastery Testing Programme. [Mastery Tests Series 4.] Tests M39-M50.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As part of a series of tests to measure mastery of specific skills in the natural sciences, copies of tests 39 through 50 include: (39) using a code; (40) naming the parts of a microscope; (41) calculating density and predicting flotation; (42) estimating metric length; (43) using SI symbols; (44) using s=vt; (45) applying a novel theory; (46)…

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. Injecting Errors for Testing Built-In Test Software

    Science.gov (United States)

    Gender, Thomas K.; Chow, James

    2010-01-01

    Two algorithms have been conceived to enable automated, thorough testing of Built-in test (BIT) software. The first algorithm applies to BIT routines that define pass/fail criteria based on values of data read from such hardware devices as memories, input ports, or registers. This algorithm simulates effects of errors in a device under test by (1) intercepting data from the device and (2) performing AND operations between the data and the data mask specific to the device. This operation yields values not expected by the BIT routine. This algorithm entails very small, permanent instrumentation of the software under test (SUT) for performing the AND operations. The second algorithm applies to BIT programs that provide services to users application programs via commands or callable interfaces and requires a capability for test-driver software to read and write the memory used in execution of the SUT. This algorithm identifies all SUT code execution addresses where errors are to be injected, then temporarily replaces the code at those addresses with small test code sequences to inject latent severe errors, then determines whether, as desired, the SUT detects the errors and recovers

  16. 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.)

  17. 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.

  18. 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.

  19. 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)

  20. 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

  1. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  2. User's manual for the G.T.M.-1 computer code

    International Nuclear Information System (INIS)

    Prado-Herrero, P.

    1992-01-01

    This document describes the GTM-1 ( Geosphere Transport Model, release-1) computer code and is intended to provide the reader with enough detailed information in order to use the code. GTM-1 was developed for the assessment of radionuclide migration by the ground water through geologic deposits whose properties can change along the pathway.GTM-1 solves the transport equation by the finite differences method ( Crank-Nicolson scheme ). It was developped for specific use within Probabilistic System Assessment (PSA) Monte Carlo Method codes; in this context the first application of GTM-1 was within the LISA (Long Term Isolation System Assessment) code. GTM-1 is also available as an independent model, which includes various submodels simulating a multi-barrier disposal system. The code has been tested with the PSACOIN ( Probabilistic System Assessment Codes intercomparison) benchmarks exercises from PSAC User Group (OECD/NEA). 10 refs., 6 Annex., 2 tabs

  3. Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

    Science.gov (United States)

    Smith, L. M.; Hochstedler, R. D.

    1997-02-01

    Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).

  4. Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

    International Nuclear Information System (INIS)

    Smith, L.M.; Hochstedler, R.D.

    1997-01-01

    Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)

  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. 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

  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. On transform coding tools under development for VP10

    Science.gov (United States)

    Parker, Sarah; Chen, Yue; Han, Jingning; Liu, Zoe; Mukherjee, Debargha; Su, Hui; Wang, Yongzhe; Bankoski, Jim; Li, Shunyao

    2016-09-01

    Google started the WebM Project in 2010 to develop open source, royaltyfree video codecs designed specifically for media on the Web. The second generation codec released by the WebM project, VP9, is currently served by YouTube, and enjoys billions of views per day. Realizing the need for even greater compression efficiency to cope with the growing demand for video on the web, the WebM team embarked on an ambitious project to develop a next edition codec, VP10, that achieves at least a generational improvement in coding efficiency over VP9. Starting from VP9, a set of new experimental coding tools have already been added to VP10 to achieve decent coding gains. Subsequently, Google joined a consortium of major tech companies called the Alliance for Open Media to jointly develop a new codec AV1. As a result, the VP10 effort is largely expected to merge with AV1. In this paper, we focus primarily on new tools in VP10 that improve coding of the prediction residue using transform coding techniques. Specifically, we describe tools that increase the flexibility of available transforms, allowing the codec to handle a more diverse range or residue structures. Results are presented on a standard test set.

  9. Creating and Testing Simulation Software

    Science.gov (United States)

    Heinich, Christina M.

    2013-01-01

    The goal of this project is to learn about the software development process, specifically the process to test and fix components of the software. The paper will cover the techniques of testing code, and the benefits of using one style of testing over another. It will also discuss the overall software design and development lifecycle, and how code testing plays an integral role in it. Coding is notorious for always needing to be debugged due to coding errors or faulty program design. Writing tests either before or during program creation that cover all aspects of the code provide a relatively easy way to locate and fix errors, which will in turn decrease the necessity to fix a program after it is released for common use. The backdrop for this paper is the Spaceport Command and Control System (SCCS) Simulation Computer Software Configuration Item (CSCI), a project whose goal is to simulate a launch using simulated models of the ground systems and the connections between them and the control room. The simulations will be used for training and to ensure that all possible outcomes and complications are prepared for before the actual launch day. The code being tested is the Programmable Logic Controller Interface (PLCIF) code, the component responsible for transferring the information from the models to the model Programmable Logic Controllers (PLCs), basic computers that are used for very simple tasks.

  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. 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.)

  12. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  14. Validation of the RALOC-mod.4 thermal-hydraulics code on evaporation transients in the Phebus containment

    International Nuclear Information System (INIS)

    Spitz, P.B.; Lemoine, F.; Tirini, S.

    1997-01-01

    IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)

  15. 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...

  16. Optical coding theory with Prime

    CERN Document Server

    Kwong, Wing C

    2013-01-01

    Although several books cover the coding theory of wireless communications and the hardware technologies and coding techniques of optical CDMA, no book has been specifically dedicated to optical coding theory-until now. Written by renowned authorities in the field, Optical Coding Theory with Prime gathers together in one volume the fundamentals and developments of optical coding theory, with a focus on families of prime codes, supplemented with several families of non-prime codes. The book also explores potential applications to coding-based optical systems and networks. Learn How to Construct

  17. 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

  18. TRAC-P validation test matrix. Revision 1.0

    International Nuclear Information System (INIS)

    Hughes, E.D.; Boyack, B.E.

    1997-01-01

    This document briefly describes the elements of the Nuclear Regulatory Commission's (NRC's) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC's software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort

  19. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  20. ETR/ITER systems code

    International Nuclear Information System (INIS)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

  1. 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

  2. Validation matrix for the assessment of thermal-hydraulic codes for VVER LOCA and transients. A report by the OECD support group on the VVER thermal-hydraulic code validation matrix

    International Nuclear Information System (INIS)

    2001-06-01

    This report deals with an internationally agreed experimental test facility matrix for the validation of best estimate thermal-hydraulic computer codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities that supplement the CSNI CCVMs and are 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. The construction of VVER Thermal-Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation Matrices (CCVM). Similar to the CCVM it is an attempt to collect together in a systematic way the best sets of available test data for VVER specific 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 in countries operating VVER reactors 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. (authors)

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. 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

  8. Tissue-specific RNA expression marks distant-acting developmental enhancers.

    Directory of Open Access Journals (Sweden)

    Han Wu

    2014-09-01

    Full Text Available Short non-coding transcripts can be transcribed from distant-acting transcriptional enhancer loci, but the prevalence of such enhancer RNAs (eRNAs within the transcriptome, and the association of eRNA expression with tissue-specific enhancer activity in vivo remain poorly understood. Here, we investigated the expression dynamics of tissue-specific non-coding RNAs in embryonic mouse tissues via deep RNA sequencing. Overall, approximately 80% of validated in vivo enhancers show tissue-specific RNA expression that correlates with tissue-specific enhancer activity. Globally, we identified thousands of tissue-specifically transcribed non-coding regions (TSTRs displaying various genomic hallmarks of bona fide enhancers. In transgenic mouse reporter assays, over half of tested TSTRs functioned as enhancers with reproducible activity in the predicted tissue. Together, our results demonstrate that tissue-specific eRNA expression is a common feature of in vivo enhancers, as well as a major source of extragenic transcription, and that eRNA expression signatures can be used to predict tissue-specific enhancers independent of known epigenomic enhancer marks.

  9. 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

  10. 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)

  11. 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...

  12. 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.

  13. RIA Fuel Codes Benchmark - Volume 1

    International Nuclear Information System (INIS)

    Marchand, Olivier; Georgenthum, Vincent; Petit, Marc; Udagawa, Yutaka; Nagase, Fumihisa; Sugiyama, Tomoyuki; Arffman, Asko; Cherubini, Marco; Dostal, Martin; Klouzal, Jan; Geelhood, Kenneth; Gorzel, Andreas; Holt, Lars; Jernkvist, Lars Olof; Khvostov, Grigori; Maertens, Dietmar; Spykman, Gerold; Nakajima, Tetsuo; Nechaeva, Olga; Panka, Istvan; Rey Gayo, Jose M.; Sagrado Garcia, Inmaculada C.; Shin, An-Dong; Sonnenburg, Heinz Guenther; Umidova, Zeynab; Zhang, Jinzhao; Voglewede, John

    2013-01-01

    Reactivity-initiated accident (RIA) fuel rod codes have been developed for a significant period of time and they all have shown their ability to reproduce some experimental results with a certain degree of adequacy. However, they sometimes rely on different specific modelling assumptions the influence of which on the final results of the calculations is difficult to evaluate. The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burnup and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including the review of experimental approaches as well as the interpretation and use of experimental data relevant for safety. As a contribution to this task, WGFS conducted a RIA code benchmark based on RIA tests performed in the Nuclear Safety Research Reactor in Tokai, Japan and tests performed or planned in CABRI reactor in Cadarache, France. Emphasis was on assessment of different modelling options for RIA fuel rod codes in terms of reproducing experimental results as well as extrapolating to typical reactor conditions. This report provides a summary of the results of this task. (authors)

  14. 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

  15. 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.

  16. TRAC-P validation test matrix. Revision 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, E.D.; Boyack, B.E.

    1997-09-05

    This document briefly describes the elements of the Nuclear Regulatory Commission`s (NRC`s) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC`s software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort.

  17. 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

  18. 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

  19. Linguistic coding deficits in foreign language learners.

    Science.gov (United States)

    Sparks, R; Ganschow, L; Pohlman, J

    1989-01-01

    As increasing numbers of colleges and universities require a foreign language for graduation in at least one of their degree programs, reports of students with difficulties in learning a second language are multiplying. Until recently, little research has been conducted to identify the nature of this problem. Recent attempts by the authors have focused upon subtle but ongoing language difficulties in these individuals as the source of their struggle to learn a foreign language. The present paper attempts to expand upon this concept by outlining a theoretical framework based upon a linguistic coding model that hypothesizes deficits in the processing of phonological, syntactic, and/or semantic information. Traditional psychoeducational assessment batteries of standardized intelligence and achievement tests generally are not sensitive to these linguistic coding deficits unless closely analyzed or, more often, used in conjunction with a more comprehensive language assessment battery. Students who have been waived from a foreign language requirement and their proposed type(s) of linguistic coding deficits are profiled. Tentative conclusions about the nature of these foreign language learning deficits are presented along with specific suggestions for tests to be used in psychoeducational evaluations.

  20. 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)

  1. 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

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

    Dogan, B.; Nikbin, K.; Petrovski, B.

    2004-01-01

    The integrity and residual life assessment of high temperature components require defects, detected or assumed to exist, through minimum allowable limits of detectable flaws using nondestructive testing methods. It relies on information obtained from the material's mechanical, uniaxial creep, creep crack initiation and growth properties. The information derived from experiments needs to be validated and harmonised following a Code of Practice that data variability between different institutions can be reduced to a minimum. The present paper reports on a Code of Practice (CoP) being prepared within the framework of the partially European Commission funded project CRETE. The novel aspect of the presented CoP is the inclusion of component relevant industrial specimen geometries. It covers testing and analysis of Creep Crack growth (CCG) in metallic materials at elevated temperature using six different cracked geometries that have been validated in. It aims to give advice on testing, measurements and analysis of creep crack growth data for a range of creep brittle to creep ductile materials using component service relevant specimen geometries and sizes. The CoP may be used for material selection criteria and inspection requirements for damage tolerant applications. In quantitative terms, these types of tests can be used to assess the individual and combined effects of metallurgical, fabrication, operating temperature, and loading conditions on creep crack growth life. Further issues will be addressed including material properties, damage and crack growth related constraint effect, stress relaxation and stress-strain fields, residual stresses, partitioning displacement, analysis of elastic creep, elastic compliance measurements

  5. 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. Technical specifications requirements: Automated reasoning applications

    International Nuclear Information System (INIS)

    Lidsky, L.M.; Dobrzeniecki, A.B.

    1990-03-01

    Several software systems were developed and tested to determine what advantages could be gained from explicitly translating complicated regulatory requirements into computerized relationships. The Technical Specifications for US nuclear power plants were chosen as the test-bed application domain, and two analysis systems were developed to monitor plant compliance with operational limits, and track and schedule equipment test and maintenance activities mandated by Technical Specifications. Choosing PROLOG as the computer language to represent these regulatory requirements resulted in a natural match between the semantic structure of the written specifications and the corollary coded rules. Additional research results affirmed the utility of declarative programming styles, explicit management of problem complexity, and attention to the robustness and flexibility of the overall software systems. 5 refs., 2 figs

  8. 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

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. Blahut-Arimoto algorithm and code design for action-dependent source coding problems

    DEFF Research Database (Denmark)

    Trillingsgaard, Kasper Fløe; Simeone, Osvaldo; Popovski, Petar

    2013-01-01

    The source coding problem with action-dependent side information at the decoder has recently been introduced to model data acquisition in resource-constrained systems. In this paper, an efficient Blahut-Arimoto-type algorithm for the numerical computation of the rate-distortion-cost function...... for this problem is proposed. Moreover, a simplified two-stage code structure based on multiplexing is put forth, whereby the first stage encodes the actions and the second stage is composed of an array of classical Wyner-Ziv codes, one for each action. Leveraging this structure, specific coding/decoding...... strategies are designed based on LDGM codes and message passing. Through numerical examples, the proposed code design is shown to achieve performance close to the rate-distortion-cost function....

  14. 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...

  15. General data analysis code for TDCR liquid scintillation counting

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, D. [Laboratorio de Metrologia de Radioisotopos, Comision Nacional de Energia Atomica, Buenos Aires (Argentina)], E-mail: drodrigu@cae.cnea.gov.ar; Arenillas, P.; Capoulat, M.E.; Balpardo, C. [Laboratorio de Metrologia de Radioisotopos, Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2008-06-15

    A non-radionuclide-specific computer code to analyze data, calculate detection efficiency and activity in a TDCR system is presented. The program was developed prioritizing flexibility in measuring conditions, parameters and calculation models. It is also intended to be well structured in order to easily replace subroutines which could eventually be improved by the user. It is written in standard FORTRAN language but a graphical interface is also available. Several tests were performed to check the ability of the code to deal with different decay schemes such as H-3, C-14, Fe-55, Mn-54 and Co-60.

  16. General data analysis code for TDCR liquid scintillation counting

    International Nuclear Information System (INIS)

    Rodrigues, D.; Arenillas, P.; Capoulat, M.E.; Balpardo, C.

    2008-01-01

    A non-radionuclide-specific computer code to analyze data, calculate detection efficiency and activity in a TDCR system is presented. The program was developed prioritizing flexibility in measuring conditions, parameters and calculation models. It is also intended to be well structured in order to easily replace subroutines which could eventually be improved by the user. It is written in standard FORTRAN language but a graphical interface is also available. Several tests were performed to check the ability of the code to deal with different decay schemes such as H-3, C-14, Fe-55, Mn-54 and Co-60

  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. Design and implementation of a scene-dependent dynamically selfadaptable wavefront coding imaging system

    Science.gov (United States)

    Carles, Guillem; Ferran, Carme; Carnicer, Artur; Bosch, Salvador

    2012-01-01

    A computational imaging system based on wavefront coding is presented. Wavefront coding provides an extension of the depth-of-field at the expense of a slight reduction of image quality. This trade-off results from the amount of coding used. By using spatial light modulators, a flexible coding is achieved which permits it to be increased or decreased as needed. In this paper a computational method is proposed for evaluating the output of a wavefront coding imaging system equipped with a spatial light modulator, with the aim of thus making it possible to implement the most suitable coding strength for a given scene. This is achieved in an unsupervised manner, thus the whole system acts as a dynamically selfadaptable imaging system. The program presented here controls the spatial light modulator and the camera, and also processes the images in a synchronised way in order to implement the dynamic system in real time. A prototype of the system was implemented in the laboratory and illustrative examples of the performance are reported in this paper. Program summaryProgram title: DynWFC (Dynamic WaveFront Coding) Catalogue identifier: AEKC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 10 483 No. of bytes in distributed program, including test data, etc.: 2 437 713 Distribution format: tar.gz Programming language: Labview 8.5 and NI Vision and MinGW C Compiler Computer: Tested on PC Intel ® Pentium ® Operating system: Tested on Windows XP Classification: 18 Nature of problem: The program implements an enhanced wavefront coding imaging system able to adapt the degree of coding to the requirements of a specific scene. The program controls the acquisition by a camera, the display of a spatial light modulator

  19. The materiality of Code

    DEFF Research Database (Denmark)

    Soon, Winnie

    2014-01-01

    This essay studies the source code of an artwork from a software studies perspective. By examining code that come close to the approach of critical code studies (Marino, 2006), I trace the network artwork, Pupufu (Lin, 2009) to understand various real-time approaches to social media platforms (MSN......, Twitter and Facebook). The focus is not to investigate the functionalities and efficiencies of the code, but to study and interpret the program level of code in order to trace the use of various technological methods such as third-party libraries and platforms’ interfaces. These are important...... to understand the socio-technical side of a changing network environment. Through the study of code, including but not limited to source code, technical specifications and other materials in relation to the artwork production, I would like to explore the materiality of code that goes beyond technical...

  20. 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.

  1. 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.)

  2. 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.…

  3. 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.

  4. 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.

  5. On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Suzuki, Motoe

    2000-01-01

    This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

  6. QR code for medical information uses.

    Science.gov (United States)

    Fontelo, Paul; Liu, Fang; Ducut, Erick G

    2008-11-06

    We developed QR code online tools, simulated and tested QR code applications for medical information uses including scanning QR code labels, URLs and authentication. Our results show possible applications for QR code in medicine.

  7. 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.

  8. 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)

  9. MORSE Monte Carlo code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described

  10. Development and application of best-estimate LWR safety analysis codes

    International Nuclear Information System (INIS)

    Reocreux, M.

    1997-01-01

    This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

  11. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    2001-01-01

    The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

  12. Purifying selection acts on coding and non-coding sequences of paralogous genes in Arabidopsis thaliana.

    Science.gov (United States)

    Hoffmann, Robert D; Palmgren, Michael

    2016-06-13

    Whole-genome duplications in the ancestors of many diverse species provided the genetic material for evolutionary novelty. Several models explain the retention of paralogous genes. However, how these models are reflected in the evolution of coding and non-coding sequences of paralogous genes is unknown. Here, we analyzed the coding and non-coding sequences of paralogous genes in Arabidopsis thaliana and compared these sequences with those of orthologous genes in Arabidopsis lyrata. Paralogs with lower expression than their duplicate had more nonsynonymous substitutions, were more likely to fractionate, and exhibited less similar expression patterns with their orthologs in the other species. Also, lower-expressed genes had greater tissue specificity. Orthologous conserved non-coding sequences in the promoters, introns, and 3' untranslated regions were less abundant at lower-expressed genes compared to their higher-expressed paralogs. A gene ontology (GO) term enrichment analysis showed that paralogs with similar expression levels were enriched in GO terms related to ribosomes, whereas paralogs with different expression levels were enriched in terms associated with stress responses. Loss of conserved non-coding sequences in one gene of a paralogous gene pair correlates with reduced expression levels that are more tissue specific. Together with increased mutation rates in the coding sequences, this suggests that similar forces of purifying selection act on coding and non-coding sequences. We propose that coding and non-coding sequences evolve concurrently following gene duplication.

  13. Auto Code Generation for Simulink-Based Attitude Determination Control System

    Science.gov (United States)

    MolinaFraticelli, Jose Carlos

    2012-01-01

    This paper details the work done to auto generate C code from a Simulink-Based Attitude Determination Control System (ADCS) to be used in target platforms. NASA Marshall Engineers have developed an ADCS Simulink simulation to be used as a component for the flight software of a satellite. This generated code can be used for carrying out Hardware in the loop testing of components for a satellite in a convenient manner with easily tunable parameters. Due to the nature of the embedded hardware components such as microcontrollers, this simulation code cannot be used directly, as it is, on the target platform and must first be converted into C code; this process is known as auto code generation. In order to generate C code from this simulation; it must be modified to follow specific standards set in place by the auto code generation process. Some of these modifications include changing certain simulation models into their atomic representations which can bring new complications into the simulation. The execution order of these models can change based on these modifications. Great care must be taken in order to maintain a working simulation that can also be used for auto code generation. After modifying the ADCS simulation for the auto code generation process, it is shown that the difference between the output data of the former and that of the latter is between acceptable bounds. Thus, it can be said that the process is a success since all the output requirements are met. Based on these results, it can be argued that this generated C code can be effectively used by any desired platform as long as it follows the specific memory requirements established in the Simulink Model.

  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. Introduction to coding and information theory

    CERN Document Server

    Roman, Steven

    1997-01-01

    This book is intended to introduce coding theory and information theory to undergraduate students of mathematics and computer science. It begins with a review of probablity theory as applied to finite sample spaces and a general introduction to the nature and types of codes. The two subsequent chapters discuss information theory: efficiency of codes, the entropy of information sources, and Shannon's Noiseless Coding Theorem. The remaining three chapters deal with coding theory: communication channels, decoding in the presence of errors, the general theory of linear codes, and such specific codes as Hamming codes, the simplex codes, and many others.

  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. Benchmark problems for radiological assessment codes. Final report

    International Nuclear Information System (INIS)

    Mills, M.; Vogt, D.; Mann, B.

    1983-09-01

    This report describes benchmark problems to test computer codes used in the radiological assessment of high-level waste repositories. The problems presented in this report will test two types of codes. The first type of code calculates the time-dependent heat generation and radionuclide inventory associated with a high-level waste package. Five problems have been specified for this code type. The second code type addressed in this report involves the calculation of radionuclide transport and dose-to-man. For these codes, a comprehensive problem and two subproblems have been designed to test the relevant capabilities of these codes for assessing a high-level waste repository setting

  18. TASS code topical report. V.1 TASS code technical manual

    International Nuclear Information System (INIS)

    Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

    1997-02-01

    TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

  19. 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.

  20. 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.

  1. 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

  2. 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)

  3. 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)

  4. The Aster code

    International Nuclear Information System (INIS)

    Delbecq, J.M.

    1999-01-01

    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.)

  5. Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

    2015-01-15

    Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

  6. Detecting non-coding selective pressure in coding regions

    Directory of Open Access Journals (Sweden)

    Blanchette Mathieu

    2007-02-01

    Full Text Available Abstract Background Comparative genomics approaches, where orthologous DNA regions are compared and inter-species conserved regions are identified, have proven extremely powerful for identifying non-coding regulatory regions located in intergenic or intronic regions. However, non-coding functional elements can also be located within coding region, as is common for exonic splicing enhancers, some transcription factor binding sites, and RNA secondary structure elements affecting mRNA stability, localization, or translation. Since these functional elements are located in regions that are themselves highly conserved because they are coding for a protein, they generally escaped detection by comparative genomics approaches. Results We introduce a comparative genomics approach for detecting non-coding functional elements located within coding regions. Codon evolution is modeled as a mixture of codon substitution models, where each component of the mixture describes the evolution of codons under a specific type of coding selective pressure. We show how to compute the posterior distribution of the entropy and parsimony scores under this null model of codon evolution. The method is applied to a set of growth hormone 1 orthologous mRNA sequences and a known exonic splicing elements is detected. The analysis of a set of CORTBP2 orthologous genes reveals a region of several hundred base pairs under strong non-coding selective pressure whose function remains unknown. Conclusion Non-coding functional elements, in particular those involved in post-transcriptional regulation, are likely to be much more prevalent than is currently known. With the numerous genome sequencing projects underway, comparative genomics approaches like that proposed here are likely to become increasingly powerful at detecting such elements.

  7. 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...

  8. Design of Packet-Based Block Codes with Shift Operators

    Directory of Open Access Journals (Sweden)

    Ilow Jacek

    2010-01-01

    Full Text Available This paper introduces packet-oriented block codes for the recovery of lost packets and the correction of an erroneous single packet. Specifically, a family of systematic codes is proposed, based on a Vandermonde matrix applied to a group of information packets to construct redundant packets, where the elements of the Vandermonde matrix are bit-level right arithmetic shift operators. The code design is applicable to packets of any size, provided that the packets within a block of information packets are of uniform length. In order to decrease the overhead associated with packet padding using shift operators, non-Vandermonde matrices are also proposed for designing packet-oriented block codes. An efficient matrix inversion procedure for the off-line design of the decoding algorithm is presented to recover lost packets. The error correction capability of the design is investigated as well. The decoding algorithm, based on syndrome decoding, to correct a single erroneous packet in a group of received packets is presented. The paper is equipped with examples of codes using different parameters. The code designs and their performance are tested using Monte Carlo simulations; the results obtained exhibit good agreement with the corresponding theoretical results.

  9. RELAP5/MOD2 code assessment

    International Nuclear Information System (INIS)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-01-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

  10. RELAP5/MOD2 code assessment

    Energy Technology Data Exchange (ETDEWEB)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-11-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

  11. 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

  12. Validation uncertainty of MATRA code for subchannel void distributions

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae-Hyun; Kim, S. J.; Kwon, H.; Seo, K. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To extend code capability to the whole core subchannel analysis, pre-conditioned Krylov matrix solvers such as BiCGSTAB and GMRES are implemented in MATRA code as well as parallel computing algorithms using MPI and OPENMP. It is coded by fortran 90, and has some user friendly features such as graphic user interface. MATRA code was approved by Korean regulation body for design calculation of integral-type PWR named SMART. The major role subchannel code is to evaluate core thermal margin through the hot channel analysis and uncertainty evaluation for CHF predictions. In addition, it is potentially used for the best estimation of core thermal hydraulic field by incorporating into multiphysics and/or multi-scale code systems. In this study we examined a validation process for the subchannel code MATRA specifically in the prediction of subchannel void distributions. The primary objective of validation is to estimate a range within which the simulation modeling error lies. The experimental data for subchannel void distributions at steady state and transient conditions was provided on the framework of OECD/NEA UAM benchmark program. The validation uncertainty of MATRA code was evaluated for a specific experimental condition by comparing the simulation result and experimental data. A validation process should be preceded by code and solution verification. However, quantification of verification uncertainty was not addressed in this study. The validation uncertainty of the MATRA code for predicting subchannel void distribution was evaluated for a single data point of void fraction measurement at a 5x5 PWR test bundle on the framework of OECD UAM benchmark program. The validation standard uncertainties were evaluated as 4.2%, 3.9%, and 2.8% with the Monte-Carlo approach at the axial levels of 2216 mm, 2669 mm, and 3177 mm, respectively. The sensitivity coefficient approach revealed similar results of uncertainties but did not account for the nonlinear effects on the

  13. 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

  14. 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.

  15. 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.)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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)

  1. SCANAIR: A transient fuel performance code

    International Nuclear Information System (INIS)

    Moal, Alain; Georgenthum, Vincent; Marchand, Olivier

    2014-01-01

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

  2. SCANAIR: A transient fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  3. On-line monitoring and inservice inspection in codes; Betriebsueberwachung und wiederkehrende Pruefungen in den Regelwerken

    Energy Technology Data Exchange (ETDEWEB)

    Bartonicek, J.; Zaiss, W. [Gemeinschaftskernkraftwerk Neckar GmbH, Neckarwestheim (Germany); Bath, H.R. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany). Geschaeftsstelle des Kerntechnischen Ausschusses (KTA)

    1999-08-01

    The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [Deutsch] Fuer die Komponentenintegritaet sind die Schaedigungsmechanismen mit dem nach den Regelwerken einzuhaltenden Abstand abzusichern. Dabei ist die jeweils vorhandene (Ist-) Qualitaet als Ausgangspunkt entscheidend. Die Absicherung der vorhandenen Qualitaet im weiteren Betrieb erfolgt durch geeignete Betriebsueberwachung und wiederkehrende Pruefungen. Die Anforderungen der Regelwerke sind vergleichbar, wobei die Bestimmung der vorhandenen Qualitaet nach einer bestimmten Betriebszeit sowie deren Absicherung im weiteren Betrieb am vollstaendigsten auf Basis des KTA-Regelwerkes moeglich ist. Die Absicherung der Komponentenintegritaet im Betrieb beruht in deutschen konventionellen Regelwerken nur auf den wiederkehrenden Pruefungen (hauptsaechlich Druckpruefungen und Sichtpruefungen). Das KTA-Regelwerk forderte hier schon immer qualifizierte

  4. Interfacial and Wall Transport Models for SPACE-CAP Code

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

    2009-01-01

    The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code

  5. Interfacial and Wall Transport Models for SPACE-CAP Code

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

  6. 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...

  7. 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...

  8. 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...

  9. Gravity inversion code

    International Nuclear Information System (INIS)

    Burkhard, N.R.

    1979-01-01

    The gravity inversion code applies stabilized linear inverse theory to determine the topography of a subsurface density anomaly from Bouguer gravity data. The gravity inversion program consists of four source codes: SEARCH, TREND, INVERT, and AVERAGE. TREND and INVERT are used iteratively to converge on a solution. SEARCH forms the input gravity data files for Nevada Test Site data. AVERAGE performs a covariance analysis on the solution. This document describes the necessary input files and the proper operation of the code. 2 figures, 2 tables

  10. Simulation of an SBLOCA Test of Shutdown Cooling System Line Break with the SMARTITL Facility using the MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yeon-Sik; Suh, Jae-Seung [System Engineering and Technology, Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sung-Uk; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    An LBLOCA (Large-Break Loss of Coolant Accident) was inherently eliminated in the design stage. The SMART design has a thermal power of 330MW. Its core exit temperature and pressurizer pressure are 323 .deg. C and 15MPa during normal operating conditions, respectively. An integral-effect test loop for SMART (SMARTITL), called FESTA (Facility for Experimental Simulation of Transients and Accidents), was designed to simulate the integral thermal-hydraulic behavior of SMART. The objectives of SMART-ITL are to investigate and understand the integral performance of reactor systems and components, and the thermal-hydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. SMART-ITL with four steam generators and PRHRS, has an advantage for a multi-loop effect compared with VISTA-ITL with a single loop. The integral-effect test data will also be used to validate the related thermal-hydraulic models of the safety analysis code such as TASS/SMR-S which is used for a performance and accident analysis of the SMART design. In addition, a scoping analysis on the scaling difference between the standard design of SMART and the basic design of SMART-ITL was performed for an SBLOCA (Small-Break Loss of Coolant Accident) scenario using a best-estimate safety analysis code, MARS-KS. This paper introduces a comparison of an SBLOCA test of a shutdown cooling system line break using SMART-ITL with its post-test calculation using the MARS-KS code. An SBLOCA test and its post-test calculation were successfully performed using the SMART-ITL facility and MARS-KS code. The SBLOCA break is a guillotine break, and its location is on the SCS line (nozzle part of the RCP suction). The steady-state conditions were achieved to satisfy the initial test conditions presented in the test requirement and its boundary conditions were properly simulated.

  11. Critical lengths of error events in convolutional codes

    DEFF Research Database (Denmark)

    Justesen, Jørn

    1994-01-01

    If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes......If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes...

  12. Critical Lengths of Error Events in Convolutional Codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Andersen, Jakob Dahl

    1998-01-01

    If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes......If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes...

  13. Verification of reactor safety codes

    International Nuclear Information System (INIS)

    Murley, T.E.

    1978-01-01

    The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

  14. 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)

  15. Pump testing - Comparison of factory vs. field test of centrifugal pumps

    International Nuclear Information System (INIS)

    Fehlau, R.

    1992-01-01

    Testing of pumps in situ, i.e., as installed in a system, will typically yield somewhat different performance results from the original manufacturer's factory test. This paper discusses some of the reasons for these variations. It shows that the factory test curves can be used for evaluation of initial acceptance tests but not for reference in normal inservice testing (IST). This is the basis for reference values used in American Society of Mechanical Engineers (ASME) Section 11 specifications and the revised ASME Code 1990

  16. 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

  17. V27 Test Report.

    Energy Technology Data Exchange (ETDEWEB)

    Stofleth, Jerome H. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Tribble, Megan Kimberly [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Crocker, Robert W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    The V27 containment vessel was procured by the US Army Recovered Chemical Material Directorate ( RCMD ) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the third EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the Code Case, is nine (9) pounds TNT - equivalent for up to 637 detonations . This report documents the results of explosive tests that were done on the vessel at Sandia National Laboratories in Albuquerque New Mexico to qualify the vessel for explosive use . The primary qualification test consisted of si x 1.5 pound charges of Composition C - 4 (equivalent to 11.25 pounds TNT) distributed around the vessel in accordance with the User Design Specification. Four subsequent tests using less explosive evaluated the effects of slight variations in orientation of the charges . All vessel acceptance criteria were met.

  18. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  19. 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).

  20. Theory of epigenetic coding.

    Science.gov (United States)

    Elder, D

    1984-06-07

    The logic of genetic control of development may be based on a binary epigenetic code. This paper revises the author's previous scheme dealing with the numerology of annelid metamerism in these terms. Certain features of the code had been deduced to be combinatorial, others not. This paradoxical contrast is resolved here by the interpretation that these features relate to different operations of the code; the combinatiorial to coding identity of units, the non-combinatorial to coding production of units. Consideration of a second paradox in the theory of epigenetic coding leads to a new solution which further provides a basis for epimorphic regeneration, and may in particular throw light on the "regeneration-duplication" phenomenon. A possible test of the model is also put forward.

  1. 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.

  2. Design of Packet-Based Block Codes with Shift Operators

    Directory of Open Access Journals (Sweden)

    Jacek Ilow

    2010-01-01

    Full Text Available This paper introduces packet-oriented block codes for the recovery of lost packets and the correction of an erroneous single packet. Specifically, a family of systematic codes is proposed, based on a Vandermonde matrix applied to a group of k information packets to construct r redundant packets, where the elements of the Vandermonde matrix are bit-level right arithmetic shift operators. The code design is applicable to packets of any size, provided that the packets within a block of k information packets are of uniform length. In order to decrease the overhead associated with packet padding using shift operators, non-Vandermonde matrices are also proposed for designing packet-oriented block codes. An efficient matrix inversion procedure for the off-line design of the decoding algorithm is presented to recover lost packets. The error correction capability of the design is investigated as well. The decoding algorithm, based on syndrome decoding, to correct a single erroneous packet in a group of n=k+r received packets is presented. The paper is equipped with examples of codes using different parameters. The code designs and their performance are tested using Monte Carlo simulations; the results obtained exhibit good agreement with the corresponding theoretical results.

  3. Utility experience in code updating of equipment built to 1974 code, Section 3, Subsection NF

    International Nuclear Information System (INIS)

    Rao, K.R.; Deshpande, N.

    1990-01-01

    This paper addresses changes to ASME Code Subsection NF and reconciles the differences between the updated codes and the as built construction code, of ASME Section III, 1974 to which several nuclear plants have been built. Since Section III is revised every three years and replacement parts complying with the construction code are invariably not available from the plant stock inventory, parts must be procured from vendors who comply with the requirements of the latest codes. Aspects of the ASME code which reflect Subsection NF are identified and compared with the later Code editions and addenda, especially up to and including the 1974 ASME code used as the basis for the plant qualification. The concern of the regulatory agencies is that if later code allowables and provisions are adopted it is possible to reduce the safety margins of the construction code. Areas of concern are highlighted and the specific changes of later codes are discerned; adoption of which, would not sacrifice the intended safety margins of the codes to which plants are licensed

  4. Random linear codes in steganography

    Directory of Open Access Journals (Sweden)

    Kamil Kaczyński

    2016-12-01

    Full Text Available Syndrome coding using linear codes is a technique that allows improvement in the steganographic algorithms parameters. The use of random linear codes gives a great flexibility in choosing the parameters of the linear code. In parallel, it offers easy generation of parity check matrix. In this paper, the modification of LSB algorithm is presented. A random linear code [8, 2] was used as a base for algorithm modification. The implementation of the proposed algorithm, along with practical evaluation of algorithms’ parameters based on the test images was made.[b]Keywords:[/b] steganography, random linear codes, RLC, LSB

  5. CATHARE code development and assessment methodologies

    International Nuclear Information System (INIS)

    Micaelli, J.C.; Barre, F.; Bestion, D.

    1995-01-01

    The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-01

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

  7. Joint Source-Channel Coding by Means of an Oversampled Filter Bank Code

    Directory of Open Access Journals (Sweden)

    Marinkovic Slavica

    2006-01-01

    Full Text Available Quantized frame expansions based on block transforms and oversampled filter banks (OFBs have been considered recently as joint source-channel codes (JSCCs for erasure and error-resilient signal transmission over noisy channels. In this paper, we consider a coding chain involving an OFB-based signal decomposition followed by scalar quantization and a variable-length code (VLC or a fixed-length code (FLC. This paper first examines the problem of channel error localization and correction in quantized OFB signal expansions. The error localization problem is treated as an -ary hypothesis testing problem. The likelihood values are derived from the joint pdf of the syndrome vectors under various hypotheses of impulse noise positions, and in a number of consecutive windows of the received samples. The error amplitudes are then estimated by solving the syndrome equations in the least-square sense. The message signal is reconstructed from the corrected received signal by a pseudoinverse receiver. We then improve the error localization procedure by introducing a per-symbol reliability information in the hypothesis testing procedure of the OFB syndrome decoder. The per-symbol reliability information is produced by the soft-input soft-output (SISO VLC/FLC decoders. This leads to the design of an iterative algorithm for joint decoding of an FLC and an OFB code. The performance of the algorithms developed is evaluated in a wavelet-based image coding system.

  8. Assessment of the computer code COBRA/CFTL

    International Nuclear Information System (INIS)

    Baxi, C.B.; Burhop, C.J.

    1981-07-01

    The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices

  9. Performance testing of real-time AI systems using the activation framework

    International Nuclear Information System (INIS)

    Becker, L.; Duckworth, J.; Laznovsky, A.; Green, P.

    1992-01-01

    This paper describes methods for automated performance testing of real-time artificial intelligence systems using the Activation Framework software development tool. The Activation Framework is suitable for applications such as the diagnosis of power system failures, which require the interpretation of large volumes of data in real-time. The Activation Framework consists of tools for compiling groups of Expert Systems rules into executable code modules, for automatically generating code modules from high level system configuration descriptions, and for automatically generating command files for program compilation and linking. It includes an operating system environment which provides the code which is common from one real-time AI applications to the next. It also includes mechanisms, described here, for automatic performance testing. The principal emphasis of this paper is on a rule based language which is used to capture performance specifications. This specification is compiled into code modules which are used to automatically test the system. This testing can validate that the system meets performance requirements during development and after maintenance. A large number of tests can be randomly generated and executed and the correctness of the outputs automatically validated. The paper also describes graph directed testing methods to minimize the number of test runs required

  10. Analysis of Semiscale Mod-1 integral test with asymmetrical break (Test S-29-1)

    International Nuclear Information System (INIS)

    Langerman, M.A.

    1977-03-01

    Selected experimental data obtained from Semiscale Mod-1 cold leg break Test S-29-1 and results obtained from analytical codes are analyzed. This test was the first integral blowdown reflood test conducted with the Mod-1 system and was a special test designed specifically to evaluate the sensitivity of the early Mod-1 core thermal response (0 to 5 sec after rupture) to the magnitude and direction of the core flow. To achieve this specific objective in Test S-29-1, the vessel side break area was reduced to approximately one-half the scaled break area associated with a 200 percent cold leg break test. The reduction in break area significantly reduced the core flow reversal that took place immediately after rupture and resulted in periods of positive core flow in the early portion of the test. The results obtained from this test are compared with results obtained from a 200 percent cold leg break test and the effect of core flow on early core thermal response is evaluated. Since Test S-29-1 was the first integral blowdown reflood test conducted with the Mod-1 system, data are also presented through the reflood stage of the test and the results are analyzed. The test data and the core thermal response calculated with the RELAP4 code are also compared

  11. TSOAK-M1: a computer code to determine tritium reaction/adsorption/release parameters from experimental results of air-detritiation tests

    International Nuclear Information System (INIS)

    Land, R.H.; Maroni, V.A.; Minkoff, M.

    1979-01-01

    A computer code has been developed which permits the determination of tritium reaction (T 2 to HTO)/adsorption/release and instrument correction parameters from enclosure (building) - detritiation test data. The code is based on a simplified model which treats each parameter as a normalized time-independent constant throughout the data-unfolding steps. Because of the complicated four-dimensional mathematical surface generated by the resulting differential equation system, occasional local-minima effects are observed, but these effects can be overcome in most instances by selecting a series of trial guesses for the initial parameter values and observing the reproducibility of final parameter values for cases where the best overall fit to experimental data is achieved. The code was then used to analyze existing small-cubicle test data with good success, and the resulting normalized parameters were employed to evaluate hypothetical reactor-building detritiation scenarios. It was concluded from the latter evaluation that the complications associated with moisture formation, adsorption, and release, particularly in terms of extended cleanup times, may not be as great as was previously thought. It is recommended that the validity of the TSOAK-M1 model be tested using data from detritiation tests conducted on large experimental enclosures (5 to 10 cm 3 ) and, if possible, actual facility buildings

  12. Code-experiment comparison on wall condensation tests in the presence of non-condensable gases-Numerical calculations for containment studies

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France); Porcheron, E.; Dumay, F.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Steam condensation on walls has been investigated in the TOSQAN vessel. Black-Right-Pointing-Pointer Experiments on 7 different tests have been performed. Black-Right-Pointing-Pointer Different steam injections and wall temperatures are used. Black-Right-Pointing-Pointer Simulations are performed in 2D using the TONUS code. Black-Right-Pointing-Pointer Code-experiments comparisons at many different locations show a good agreement. - Abstract: During the course of a severe Pressurized Water Reactor accident, pressurization of the containment occurs and hydrogen can be produced by the reactor core oxidation and distributed in the containment according to convection flows and wall condensation. Filmwise wall condensation in the presence of non-condensable gases is a subject of many interests and extensive studies have been performed in the past. Some empirical correlations have demonstrated their limit for extrapolation under different thermal-hydraulic conditions and at different geometries/scales. The French Institute for Radiological Protection and Nuclear Safety (IRSN) has developed a numerical tool and an experimental facility in order to investigate free convection flows in the presence of condensation. The objective of this paper is to present numerical results obtained on different wall condensation tests in 7 m{sup 3} volume vessel (TOSQAN facility), and to compare them with the experimental ones. Over eight tests are considered here, and code-experiment comparison is performed on many different locations, giving an extensive insight of the code assessment for air-steam mixture flows involving wall condensation in the presence of non-condensable gases.

  13. Unit Testing for the Application Control Language (ACL) Software

    Science.gov (United States)

    Heinich, Christina Marie

    2014-01-01

    In the software development process, code needs to be tested before it can be packaged for release in order to make sure the program actually does what it says is supposed to happen as well as to check how the program deals with errors and edge cases (such as negative or very large numbers). One of the major parts of the testing process is unit testing, where you test specific units of the code to make sure each individual part of the code works. This project is about unit testing many different components of the ACL software and fixing any errors encountered. To do this, mocks of other objects need to be created and every line of code needs to be exercised to make sure every case is accounted for. Mocks are important to make because it gives direct control of the environment the unit lives in instead of attempting to work with the entire program. This makes it easier to achieve the second goal of exercising every line of code.

  14. 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.

  15. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the System Design Description (SDD) for the TMAD Code System, which includes the TMAD code and the LIBMAKR code. The SDD provides a detailed description of the theory behind the code, and the implementation of that theory. It is essential for anyone who is attempting to review or modify the code or who otherwise needs to understand the internal workings of the code. In addition, this document includes, in Appendix A, the System Requirements Specification for the TMAD System

  16. Code Disentanglement: Initial Plan

    Energy Technology Data Exchange (ETDEWEB)

    Wohlbier, John Greaton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Timothy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rockefeller, Gabriel M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Calef, Matthew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-27

    The first step to making more ambitious changes in the EAP code base is to disentangle the code into a set of independent, levelized packages. We define a package as a collection of code, most often across a set of files, that provides a defined set of functionality; a package a) can be built and tested as an entity and b) fits within an overall levelization design. Each package contributes one or more libraries, or an application that uses the other libraries. A package set is levelized if the relationships between packages form a directed, acyclic graph and each package uses only packages at lower levels of the diagram (in Fortran this relationship is often describable by the use relationship between modules). Independent packages permit independent- and therefore parallel|development. The packages form separable units for the purposes of development and testing. This is a proven path for enabling finer-grained changes to a complex code.

  17. Testing Scientific Software: A Systematic Literature Review

    Science.gov (United States)

    Kanewala, Upulee; Bieman, James M.

    2014-01-01

    Context Scientific software plays an important role in critical decision making, for example making weather predictions based on climate models, and computation of evidence for research publications. Recently, scientists have had to retract publications due to errors caused by software faults. Systematic testing can identify such faults in code. Objective This study aims to identify specific challenges, proposed solutions, and unsolved problems faced when testing scientific software. Method We conducted a systematic literature survey to identify and analyze relevant literature. We identified 62 studies that provided relevant information about testing scientific software. Results We found that challenges faced when testing scientific software fall into two main categories: (1) testing challenges that occur due to characteristics of scientific software such as oracle problems and (2) testing challenges that occur due to cultural differences between scientists and the software engineering community such as viewing the code and the model that it implements as inseparable entities. In addition, we identified methods to potentially overcome these challenges and their limitations. Finally we describe unsolved challenges and how software engineering researchers and practitioners can help to overcome them. Conclusions Scientific software presents special challenges for testing. Specifically, cultural differences between scientist developers and software engineers, along with the characteristics of the scientific software make testing more difficult. Existing techniques such as code clone detection can help to improve the testing process. Software engineers should consider special challenges posed by scientific software such as oracle problems when developing testing techniques. PMID:25125798

  18. Testing Scientific Software: A Systematic Literature Review.

    Science.gov (United States)

    Kanewala, Upulee; Bieman, James M

    2014-10-01

    Scientific software plays an important role in critical decision making, for example making weather predictions based on climate models, and computation of evidence for research publications. Recently, scientists have had to retract publications due to errors caused by software faults. Systematic testing can identify such faults in code. This study aims to identify specific challenges, proposed solutions, and unsolved problems faced when testing scientific software. We conducted a systematic literature survey to identify and analyze relevant literature. We identified 62 studies that provided relevant information about testing scientific software. We found that challenges faced when testing scientific software fall into two main categories: (1) testing challenges that occur due to characteristics of scientific software such as oracle problems and (2) testing challenges that occur due to cultural differences between scientists and the software engineering community such as viewing the code and the model that it implements as inseparable entities. In addition, we identified methods to potentially overcome these challenges and their limitations. Finally we describe unsolved challenges and how software engineering researchers and practitioners can help to overcome them. Scientific software presents special challenges for testing. Specifically, cultural differences between scientist developers and software engineers, along with the characteristics of the scientific software make testing more difficult. Existing techniques such as code clone detection can help to improve the testing process. Software engineers should consider special challenges posed by scientific software such as oracle problems when developing testing techniques.

  19. Summary - COG: A new point-wise Monte Carlo code for burnup credit analysis

    International Nuclear Information System (INIS)

    Alesso, H.P.

    1989-01-01

    COG, a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL) for the Cray-1, solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) other particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems and a wide variety of criticality problems. COG is similar to a number of other computer codes used in the shielding community. Each code is a little different in its geometry input and its random-walk modification options. COG is a Monte Carlo code specifically designed for the CRAY (in 1986) to be as precise as the current state of physics knowledge. It has been extensively benchmarked and used as a shielding code at LLNL since 1986, and has recently been extended to accomplish criticality calculations. It will make an excellent tool for future shipping cask studies

  20. The nuclear codes and guidelines

    International Nuclear Information System (INIS)

    Sonter, M.

    1984-01-01

    This paper considers problems faced by the mining industry when implementing the nuclear codes of practice. Errors of interpretation are likely. A major criticism is that the guidelines to the codes must be seen as recommendations only. They are not regulations. Specific clauses in the guidelines are criticised

  1. Code-Mixing and Code Switchingin The Process of Learning

    Directory of Open Access Journals (Sweden)

    Diyah Atiek Mustikawati

    2016-09-01

    Full Text Available This study aimed to describe a form of code switching and code mixing specific form found in the teaching and learning activities in the classroom as well as determining factors influencing events stand out that form of code switching and code mixing in question.Form of this research is descriptive qualitative case study which took place in Al Mawaddah Boarding School Ponorogo. Based on the analysis and discussion that has been stated in the previous chapter that the form of code mixing and code switching learning activities in Al Mawaddah Boarding School is in between the use of either language Java language, Arabic, English and Indonesian, on the use of insertion of words, phrases, idioms, use of nouns, adjectives, clauses, and sentences. Code mixing deciding factor in the learning process include: Identification of the role, the desire to explain and interpret, sourced from the original language and its variations, is sourced from a foreign language. While deciding factor in the learning process of code, includes: speakers (O1, partners speakers (O2, the presence of a third person (O3, the topic of conversation, evoke a sense of humour, and just prestige. The significance of this study is to allow readers to see the use of language in a multilingual society, especially in AL Mawaddah boarding school about the rules and characteristics variation in the language of teaching and learning activities in the classroom. Furthermore, the results of this research will provide input to the ustadz / ustadzah and students in developing oral communication skills and the effectiveness of teaching and learning strategies in boarding schools.

  2. 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)

  3. 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)

  4. Benchmarking the Multidimensional Stellar Implicit Code MUSIC

    Science.gov (United States)

    Goffrey, T.; Pratt, J.; Viallet, M.; Baraffe, I.; Popov, M. V.; Walder, R.; Folini, D.; Geroux, C.; Constantino, T.

    2017-04-01

    We present the results of a numerical benchmark study for the MUltidimensional Stellar Implicit Code (MUSIC) based on widely applicable two- and three-dimensional compressible hydrodynamics problems relevant to stellar interiors. MUSIC is an implicit large eddy simulation code that uses implicit time integration, implemented as a Jacobian-free Newton Krylov method. A physics based preconditioning technique which can be adjusted to target varying physics is used to improve the performance of the solver. The problems used for this benchmark study include the Rayleigh-Taylor and Kelvin-Helmholtz instabilities, and the decay of the Taylor-Green vortex. Additionally we show a test of hydrostatic equilibrium, in a stellar environment which is dominated by radiative effects. In this setting the flexibility of the preconditioning technique is demonstrated. This work aims to bridge the gap between the hydrodynamic test problems typically used during development of numerical methods and the complex flows of stellar interiors. A series of multidimensional tests were performed and analysed. Each of these test cases was analysed with a simple, scalar diagnostic, with the aim of enabling direct code comparisons. As the tests performed do not have analytic solutions, we verify MUSIC by comparing it to established codes including ATHENA and the PENCIL code. MUSIC is able to both reproduce behaviour from established and widely-used codes as well as results expected from theoretical predictions. This benchmarking study concludes a series of papers describing the development of the MUSIC code and provides confidence in future applications.

  5. Hydrogen burn assessment with the CONTAIN code

    International Nuclear Information System (INIS)

    van Rij, H.M.

    1986-01-01

    The CONTAIN computer code was developed at Sandia National Laboratories, under contract to the US Nuclear Regulatory Commission (NRC). The code is intended for calculations of containment loads during severe accidents and for prediction of the radioactive source term in the event that the containment leaks or fails. A strong point of the CONTAIN code is the continuous interaction of the thermal-hydraulics phenomena, aerosol behavior and fission product behavior. The CONTAIN code can be used for Light Water Reactors as well as Liquid Metal Reactors. In order to evaluate the CONTAIN code on its merits, comparisons between the code and experiments must be made. In this paper, CONTAIN calculations for the hydrogen burn experiments, carried out at the Nevada Test Site (NTS), are presented and compared with the experimental data. In the Large-Scale Hydrogen Combustion Facility at the NTS, 21 tests have been carried out. These tests were sponsored by the NRC and the Electric Power Research Institute (EPRI). The tests, carried out by EG and G, were performed in a spherical vessel 16 m in diameter with a design pressure of 700 kPa, substantially higher than that of most commercial nuclear containment buildings

  6. 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

  7. 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.

  8. The structure of affective action representations: temporal binding of affective response codes.

    Science.gov (United States)

    Eder, Andreas B; Müsseler, Jochen; Hommel, Bernhard

    2012-01-01

    Two experiments examined the hypothesis that preparing an action with a specific affective connotation involves the binding of this action to an affective code reflecting this connotation. This integration into an action plan should lead to a temporary occupation of the affective code, which should impair the concurrent representation of affectively congruent events, such as the planning of another action with the same valence. This hypothesis was tested with a dual-task setup that required a speeded choice between approach- and avoidance-type lever movements after having planned and before having executed an evaluative button press. In line with the code-occupation hypothesis, slower lever movements were observed when the lever movement was affectively compatible with the prepared evaluative button press than when the two actions were affectively incompatible. Lever movements related to approach and avoidance and evaluative button presses thus seem to share a code that represents affective meaning. A model of affective action control that is based on the theory of event coding is discussed.

  9. 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...

  10. 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.

  11. ComboCoding: Combined intra-/inter-flow network coding for TCP over disruptive MANETs

    Directory of Open Access Journals (Sweden)

    Chien-Chia Chen

    2011-07-01

    Full Text Available TCP over wireless networks is challenging due to random losses and ACK interference. Although network coding schemes have been proposed to improve TCP robustness against extreme random losses, a critical problem still remains of DATA–ACK interference. To address this issue, we use inter-flow coding between DATA and ACK to reduce the number of transmissions among nodes. In addition, we also utilize a “pipeline” random linear coding scheme with adaptive redundancy to overcome high packet loss over unreliable links. The resulting coding scheme, ComboCoding, combines intra-flow and inter-flow coding to provide robust TCP transmission in disruptive wireless networks. The main contributions of our scheme are twofold; the efficient combination of random linear coding and XOR coding on bi-directional streams (DATA and ACK, and the novel redundancy control scheme that adapts to time-varying and space-varying link loss. The adaptive ComboCoding was tested on a variable hop string topology with unstable links and on a multipath MANET with dynamic topology. Simulation results show that TCP with ComboCoding delivers higher throughput than with other coding options in high loss and mobile scenarios, while introducing minimal overhead in normal operation.

  12. Integrated Validation System for a Thermal-hydraulic System Code, TASS/SMR-S

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee-Kyung; Kim, Hyungjun; Kim, Soo Hyoung; Hwang, Young-Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Hyeon-Soo [Chungnam National University, Daejeon (Korea, Republic of)

    2015-10-15

    Development including enhancement and modification of thermal-hydraulic system computer code is indispensable to a new reactor, SMART. Usually, a thermal-hydraulic system code validation is achieved by a comparison with the results of corresponding physical effect tests. In the reactor safety field, a similar concept, referred to as separate effect tests has been used for a long time. But there are so many test data for comparison because a lot of separate effect tests and integral effect tests are required for a code validation. It is not easy to a code developer to validate a computer code whenever a code modification is occurred. IVS produces graphs which shown the comparison the code calculation results with the corresponding test results automatically. IVS was developed for a validation of TASS/SMR-S code. The code validation could be achieved by a comparison code calculation results with corresponding test results. This comparison was represented as a graph for convenience. IVS is useful before release a new code version. The code developer can validate code result easily using IVS. Even during code development, IVS could be used for validation of code modification. The code developer could gain a confidence about his code modification easily and fast and could be free from tedious and long validation work. The popular software introduced in IVS supplies better usability and portability.

  13. SPEAR-FCODE-GAMMA functional specifications. Final report

    International Nuclear Information System (INIS)

    Fiero, I.B.

    1983-03-01

    SPEAR FCODE GAMMA (SFG), a conceptual fuel-performance code for use in licensing analyses, has been defined and characterized as a set of functional specifications. The potential licensing-related applications of SFG are established and discussed. General code specifications including regulatory, interface, hardware application, code model and software, and operational specifications are discussed. The code input and output information including data requirements as well as formatting aspects are detailed. Finally, the SFG code-accuracy guidelines are established and the validation process is described

  14. 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

  15. 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.

  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. The HELIOS-2 lattice physics code

    International Nuclear Information System (INIS)

    Wemple, C.A.; Gheorghiu, H-N.M.; Stamm'ler, R.J.J.; Villarino, E.A.

    2008-01-01

    Major advances have been made in the HELIOS code, resulting in the impending release of a new version, HELIOS-2. The new code includes a method of characteristics (MOC) transport solver to supplement the existing collision probabilities (CP) solver. A 177-group, ENDF/B-VII nuclear data library has been developed for inclusion with the new code package. Computational tests have been performed to verify the performance of the MOC solver against the CP solver, and validation testing against computational and measured benchmarks is underway. Results to-date of the verification and validation testing are presented, demonstrating the excellent performance of the new transport solver and nuclear data library. (Author)

  18. Certification plan for safety and PRA codes

    International Nuclear Information System (INIS)

    Toffer, H.; Crowe, R.D.; Ades, M.J.

    1990-05-01

    A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA ampersand PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations

  19. Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base

    Science.gov (United States)

    Savage, B.; Snoke, J. A.

    2017-12-01

    The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years

  20. Lattice Index Coding

    OpenAIRE

    Natarajan, Lakshmi; Hong, Yi; Viterbo, Emanuele

    2014-01-01

    The index coding problem involves a sender with K messages to be transmitted across a broadcast channel, and a set of receivers each of which demands a subset of the K messages while having prior knowledge of a different subset as side information. We consider the specific case of noisy index coding where the broadcast channel is Gaussian and every receiver demands all the messages from the source. Instances of this communication problem arise in wireless relay networks, sensor networks, and ...

  1. Analysis of Peach Bottom turbine trip tests

    International Nuclear Information System (INIS)

    Cheng, H.S.; Lu, M.S.; Hsu, C.J.; Shier, W.G.; Diamond, D.J.; Levine, M.M.; Odar, F.

    1979-01-01

    Current interest in the analysis of turbine trip transients has been generated by the recent tests performed at the Peach Bottom (Unit 2) reactor. Three tests, simulating turbine trip transients, were performed at different initial power and coolant flow conditions. The data from these tests provide considerable information to aid qualification of computer codes that are currently used in BWR design analysis. The results are presented of an analysis of a turbine trip transient using the RELAP-3B and the BNL-TWIGL computer codes. Specific results are provided comparing the calculated reactor power and system pressures with the test data. Excellent agreement for all three test transients is evident from the comparisons

  2. System verification and validation report for the TMAD code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the Verification and Validation Report for the TMAD code system, which includes the TMAD code and the LIBMAKR Code. The TMAD code was commissioned to facilitate the interpretation of moisture probe measurements in the Hanford Site waste tanks. In principle, the code is an interpolation routine that acts over a library of benchmark data based on two independent variables, typically anomaly size and moisture content. Two additional variables, anomaly type and detector type, can also be considered independent variables, but no interpolation is done over them. The dependent variable is detector response. The intent is to provide the code with measured detector responses from two or more detectors. The code will then interrogate (and interpolate upon) the benchmark data library and find the anomaly-type/anomaly-size/moisture-content combination that provides the closest match to the measured data. The primary purpose of this document is to provide the results of the system testing and the conclusions based thereon. The results of the testing process are documented in the body of the report. Appendix A gives the test plan, including test procedures, used in conducting the tests. Appendix B lists the input data required to conduct the tests, and Appendices C and 0 list the numerical results of the tests

  3. 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.

  4. 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.

  5. Automated Test Case Generation

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    I would like to present the concept of automated test case generation. I work on it as part of my PhD and I think it would be interesting also for other people. It is also the topic of a workshop paper that I am introducing in Paris. (abstract below) Please note that the talk itself would be more general and not about the specifics of my PhD, but about the broad field of Automated Test Case Generation. I would introduce the main approaches (combinatorial testing, symbolic execution, adaptive random testing) and their advantages and problems. (oracle problem, combinatorial explosion, ...) Abstract of the paper: Over the last decade code-based test case generation techniques such as combinatorial testing or dynamic symbolic execution have seen growing research popularity. Most algorithms and tool implementations are based on finding assignments for input parameter values in order to maximise the execution branch coverage. Only few of them consider dependencies from outside the Code Under Test’s scope such...

  6. First analysis of AGS0, LT2 and E9 CABRI tests with the new SFR safety code ASTEC-Na

    International Nuclear Information System (INIS)

    Perez-Martin, Sara; Bandini, Giacomino; Matuzas, Vaidas; Buck, Michael; Girault, Nathalie

    2015-01-01

    Within the framework of the European JASMIN project, the ASTEC-Na code is being developed for safety analysis of severe accidents in SFR. In the first phase of validation of the ASTEC-Na fuel thermo-mechanical models three in-pile tests conducted in the CABRI experimental reactor have been selected to be analysed. We present here the preliminary results of the simulation of two Transient Over Power tests and one power ramp test (AGS0, LT2 and E9, respectively) where no pin failure occurred during the transient. We present the comparison of ASTEC-Na results against experimental data and other safety code results for the initial steady state conditions prior to the transient onset as well as for the fuel pin behaviour during the transients. (author)

  7. Fundamentals of information theory and coding design

    CERN Document Server

    Togneri, Roberto

    2003-01-01

    In a clear, concise, and modular format, this book introduces the fundamental concepts and mathematics of information and coding theory. The authors emphasize how a code is designed and discuss the main properties and characteristics of different coding algorithms along with strategies for selecting the appropriate codes to meet specific requirements. They provide comprehensive coverage of source and channel coding, address arithmetic, BCH, and Reed-Solomon codes and explore some more advanced topics such as PPM compression and turbo codes. Worked examples and sets of basic and advanced exercises in each chapter reinforce the text's clear explanations of all concepts and methodologies.

  8. Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

  9. Verification and benchmarking of PORFLO: an equivalent porous continuum code for repository scale analysis

    International Nuclear Information System (INIS)

    Eyler, L.L.; Budden, M.J.

    1984-11-01

    The objective of this work was to perform an assessment of prediction capabilities and features of the PORFLO code in relation to its intended use in the Basalt Waste Isolation Project. This objective was to be accomplished through a code verification and benchmarking task. Results were to be documented which either support correctness of prediction capabilities or identify areas of intended application in which the code exhibits weaknesses. A test problem set consisting of 10 problems was developed. Results of PORFLO simulations of these problems were provided for use in this work. The 10 problems were designed to test the three basic computational capabilities or categories of the code. Broken down by physical process, these are heat transfer, fluid flow, and radionuclide transport. Two verification problems were included within each of these categories. They were problems designed to test basic features of PORFLO for which analytical solutions are available for use as a known comparison basis. Hence they are referred to as verification problems. Of the remaining four problems, one repository scale problem representative of intended PORFLO use within BWIP was included in each of the three basic capabilities categories. The remaining problem was a case specifically designed to test features of decay and retardation in radionuclide transport. These four problems are referred to as benchmarking problems, because results computed with an additional computer code were used as a basis for comparison. 38 figures

  10. Inelasticity and the ASME code

    International Nuclear Information System (INIS)

    Berman, I.

    1978-01-01

    Although it may have more general applicability, this paper is specifically concerned with some aspects of plasticity for class I nuclear components that are contained in section III of the ASME Boiler and Pressure Vessel Code. It directly addresses design for components at temperatures at which creep is not a factor. A review is made of the relationship of plasticity to each of the three failure modes that the stress limits are intended to prevent. It is found that the prevention of bursting and gross distortion from a single application of pressure and the prevention of fatigue failure caused by repeated cycles of peak stresses are well supported by experimental results. The experimental verification for the rules to show that the primary plus secondary stresses shakedown to elastic behavior is not clear. Various directed efforts which could lead to greater assurance of shakedown to elastic behavior are suggested. The major approach should be a massive program to develop a test matrix of experimental information for various portions of each component of interest in the Code. (Auth.)

  11. Temporal code-driven stimulation: definition and application to electric fish signaling

    Directory of Open Access Journals (Sweden)

    Angel Lareo

    2016-10-01

    Full Text Available Closed-loop activity-dependent stimulation is a powerful methodology to assess information processing in biological systems. In this context, the development of novel protocols, their implementation in bioinformatics toolboxes and their application to different description levels open up a wide range of possibilities in the study of biological systems. We developed a methodology for studying biological signals representing them as temporal sequences of binary events. A specific sequence of these events (code is chosen to deliver a predefined stimulation in a closed-loop manner. The response to this code-driven stimulation can be used to characterize the system. This methodology was implemented in a real time toolbox and tested in the context of electric fish signaling. We show that while there are codes that evoke a response that cannot be distinguished from a control recording without stimulation, other codes evoke a characteristic distinct response. We also compare the code-driven response to open-loop stimulation. The discussed experiments validate the proposed methodology and the software toolbox.

  12. Guidelines for inservice testing at nuclear power plants

    International Nuclear Information System (INIS)

    Campbell, P.

    1995-04-01

    The staff of the U.S. Nuclear Regulatory Commission (NRC) gives licensees guidelines and recommendations for developing and implementing programs for the inservice testing of pumps and valves at commercial nuclear power plants. The staff discusses the regulations; the components to be included in an inservice testing program; and the preparation and content of cold shutdown justifications, refueling outage justifications, and requests for relief from the American Society of Mechanical Engineers Code requirements. The staff also gives specific guidance on relief acceptable to the NRC and advises licensees in the use of this information at their facilities. The staff discusses the revised standard technical specifications for the inservice testing program requirements and gives guidance on the process a licensee may follow upon finding an instance of noncompliance with the Code

  13. Rulemaking efforts on codes and standards

    International Nuclear Information System (INIS)

    Millman, G.C.

    1992-01-01

    Section 50.55a of the NRC regulations provides a mechanism for incorporating national codes and standards into the regulatory process. It incorporates by reference ASME Boiler and Pressure Vessel Code (ASME B and PV Code) Section 3 rules for construction and Section 11 rules for inservice inspection and inservice testing. The regulation is periodically amended to update these references. The rulemaking process, as applied to Section 50.55a amendments, is overviewed to familiarize users with associated internal activities of the NRC staff and the manner in which public comments are integrated into the process. The four ongoing rulemaking actions that would individually amend Section 50.55a are summarized. Two of the actions would directly impact requirements for inservice testing. Benefits accrued with NRC endorsement of the ASME B and PV Code, and possible future endorsement of the ASME Operations and Maintenance Code (ASME OM Code), are identified. Emphasis is placed on the need for code writing committees to be especially sensitive to user feedback on code rules incorporated into the regulatory process to ensure that the rules are complete, technically accurate, clear, practical, and enforceable

  14. 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)

  15. 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.

  16. An UI Layout Files Analyzer for Test Data Generation

    Directory of Open Access Journals (Sweden)

    Paul POCATILU

    2014-01-01

    Full Text Available Prevention actions (trainings, audits and inspections (tests, validations, code reviews are the crucial factors in achieving a high quality level for any software application simply because low investments in this area are leading to significant expenses in terms of corrective actions needed for defect fixing. Mobile applications testing involves the use of various tools and scenarios. An important process is represented by test data generation. This paper proposes a test data generator (TDG system for mobile applications using several sources for test data and it focuses on the UI layout files analyzer module. The proposed architecture aims to reduce time-to-market for mobile applications. The focus is on test data generators based on the source code, user interface layout files (using markup languages like XML or XAML and application specifications. In order to assure a common interface for test data generators, an XML or JSON-based language called Data Specification Language (DSL is proposed.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. Coding theory and cryptography the essentials

    CERN Document Server

    Hankerson, DC; Leonard, DA; Phelps, KT; Rodger, CA; Wall, JR; Wall, J R

    2000-01-01

    Containing data on number theory, encryption schemes, and cyclic codes, this highly successful textbook, proven by the authors in a popular two-quarter course, presents coding theory, construction, encoding, and decoding of specific code families in an ""easy-to-use"" manner appropriate for students with only a basic background in mathematics offering revised and updated material on the Berlekamp-Massey decoding algorithm and convolutional codes. Introducing the mathematics as it is needed and providing exercises with solutions, this edition includes an extensive section on cryptography, desig

  2. Schroedinger’s Code: A Preliminary Study on Research Source Code Availability and Link Persistence in Astrophysics

    Science.gov (United States)

    Allen, Alice; Teuben, Peter J.; Ryan, P. Wesley

    2018-05-01

    We examined software usage in a sample set of astrophysics research articles published in 2015 and searched for the source codes for the software mentioned in these research papers. We categorized the software to indicate whether the source code is available for download and whether there are restrictions to accessing it, and if the source code is not available, whether some other form of the software, such as a binary, is. We also extracted hyperlinks from one journal’s 2015 research articles, as links in articles can serve as an acknowledgment of software use and lead to the data used in the research, and tested them to determine which of these URLs are still accessible. For our sample of 715 software instances in the 166 articles we examined, we were able to categorize 418 records as according to whether source code was available and found that 285 unique codes were used, 58% of which offered the source code for download. Of the 2558 hyperlinks extracted from 1669 research articles, at best, 90% of them were available over our testing period.

  3. Improving the accuracy of operation coding in surgical discharge summaries

    Science.gov (United States)

    Martinou, Eirini; Shouls, Genevieve; Betambeau, Nadine

    2014-01-01

    Procedural coding in surgical discharge summaries is extremely important; as well as communicating to healthcare staff which procedures have been performed, it also provides information that is used by the hospital's coding department. The OPCS code (Office of Population, Censuses and Surveys Classification of Surgical Operations and Procedures) is used to generate the tariff that allows the hospital to be reimbursed for the procedure. We felt that the OPCS coding on discharge summaries was often incorrect within our breast and endocrine surgery department. A baseline measurement over two months demonstrated that 32% of operations had been incorrectly coded, resulting in an incorrect tariff being applied and an estimated loss to the Trust of £17,000. We developed a simple but specific OPCS coding table in collaboration with the clinical coding team and breast surgeons that summarised all operations performed within our department. This table was disseminated across the team, specifically to the junior doctors who most frequently complete the discharge summaries. Re-audit showed 100% of operations were accurately coded, demonstrating the effectiveness of the coding table. We suggest that specifically designed coding tables be introduced across each surgical department to ensure accurate OPCS codes are used to produce better quality surgical discharge summaries and to ensure correct reimbursement to the Trust. PMID:26734286

  4. 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

  5. Elements of algebraic coding systems

    CERN Document Server

    Cardoso da Rocha, Jr, Valdemar

    2014-01-01

    Elements of Algebraic Coding Systems is an introductory text to algebraic coding theory. In the first chapter, you'll gain inside knowledge of coding fundamentals, which is essential for a deeper understanding of state-of-the-art coding systems. This book is a quick reference for those who are unfamiliar with this topic, as well as for use with specific applications such as cryptography and communication. Linear error-correcting block codes through elementary principles span eleven chapters of the text. Cyclic codes, some finite field algebra, Goppa codes, algebraic decoding algorithms, and applications in public-key cryptography and secret-key cryptography are discussed, including problems and solutions at the end of each chapter. Three appendices cover the Gilbert bound and some related derivations, a derivation of the Mac- Williams' identities based on the probability of undetected error, and two important tools for algebraic decoding-namely, the finite field Fourier transform and the Euclidean algorithm f...

  6. 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

  7. Test Particle Simulations of Electron Injection by the Bursty Bulk Flows (BBFs) using High Resolution Lyon-Feddor-Mobarry (LFM) Code

    Science.gov (United States)

    Eshetu, W. W.; Lyon, J.; Wiltberger, M. J.; Hudson, M. K.

    2017-12-01

    Test particle simulations of electron injection by the bursty bulk flows (BBFs) have been done using a test particle tracer code [1], and the output fields of the Lyon-Feddor-Mobarry global magnetohydro- dynamics (MHD) code[2]. The MHD code was run with high resolu- tion (oct resolution), and with specified solar wind conditions so as to reproduce the observed qualitative picture of the BBFs [3]. Test par- ticles were injected so that they interact with earthward propagating BBFs. The result of the simulation shows that electrons are pushed ahead of the BBFs and accelerated into the inner magnetosphere. Once electrons are in the inner magnetosphere they are further energized by drift resonance with the azimuthal electric field. In addition pitch angle scattering of electrons resulting in the violation conservation of the first adiabatic invariant has been observed. The violation of the first adiabatic invariant occurs as electrons cross a weak magnetic field region with a strong gradient of the field perturbed by the BBFs. References 1. Kress, B. T., Hudson,M. K., Looper, M. D. , Albert, J., Lyon, J. G., and Goodrich, C. C. (2007), Global MHD test particle simulations of ¿ 10 MeV radiation belt electrons during storm sudden commencement, J. Geophys. Res., 112, A09215, doi:10.1029/2006JA012218. Lyon,J. G., Fedder, J. A., and Mobarry, C.M., The Lyon- Fedder-Mobarry (LFM) Global MHD Magnetospheric Simulation Code (2004), J. Atm. And Solar-Terrestrial Phys., 66, Issue 15-16, 1333- 1350,doi:10.1016/j.jastp. Wiltberger, Merkin, M., Lyon, J. G., and Ohtani, S. (2015), High-resolution global magnetohydrodynamic simulation of bursty bulk flows, J. Geophys. Res. Space Physics, 120, 45554566, doi:10.1002/2015JA021080.

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-15

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

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

    International Nuclear Information System (INIS)

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

    2012-07-01

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

  11. Governance codes: facts or fictions? a study of governance codes in colombia1,2

    Directory of Open Access Journals (Sweden)

    Julián Benavides Franco

    2010-10-01

    Full Text Available This article studies the effects on accounting performance and financing decisions of Colombian firms after issuing a corporate governance code. We assemble a database of Colombian issuers and test the hypotheses of improved performance and higher leverage after issuing a code. The results show that the firms’ return on assets after the code introduction improves in excess of 1%; the effect is amplified by the code quality. Additionally, the firms leverage increased, in excess of 5%, when the code quality was factored into the analysis. These results suggest that controlling parties commitment to self restrain, by reducing their private benefits and/or the expropriation of non controlling parties, through the code introduction, is indeed an effective measure and that the financial markets agree, increasing the supply of funds to the firms.

  12. Methodology, status and plans for development and assessment of Cathare code

    Energy Technology Data Exchange (ETDEWEB)

    Bestion, D.; Barre, F.; Faydide, B. [CEA - Grenoble (France)

    1997-07-01

    This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests or integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.

  13. 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

  14. 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

  15. On Coding Non-Contiguous Letter Combinations

    Directory of Open Access Journals (Sweden)

    Frédéric eDandurand

    2011-06-01

    Full Text Available Starting from the hypothesis that printed word identification initially involves the parallel mapping of visual features onto location-specific letter identities, we analyze the type of information that would be involved in optimally mapping this location-specific orthographic code onto a location-invariant lexical code. We assume that some intermediate level of coding exists between individual letters and whole words, and that this involves the representation of letter combinations. We then investigate the nature of this intermediate level of coding given the constraints of optimality. This intermediate level of coding is expected to compress data while retaining as much information as possible about word identity. Information conveyed by letters is a function of how much they constrain word identity and how visible they are. Optimization of this coding is a combination of minimizing resources (using the most compact representations and maximizing information. We show that in a large proportion of cases, non-contiguous letter sequences contain more information than contiguous sequences, while at the same time requiring less precise coding. Moreover, we found that the best predictor of human performance in orthographic priming experiments was within-word ranking of conditional probabilities, rather than average conditional probabilities. We conclude that from an optimality perspective, readers learn to select certain contiguous and non-contiguous letter combinations as information that provides the best cue to word identity.

  16. Energy information data base: report number codes

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-09-01

    Each report processed by the US DOE Technical Information Center is identified by a unique report number consisting of a code plus a sequential number. In most cases, the code identifies the originating installation. In some cases, it identifies a specific program or a type of publication. Listed in this publication are all codes that have been used by DOE in cataloging reports. This compilation consists of two parts. Part I is an alphabetical listing of report codes identified with the issuing installations that have used the codes. Part II is an alphabetical listing of installations identified with codes each has used. (RWR)

  17. Energy information data base: report number codes

    International Nuclear Information System (INIS)

    1979-09-01

    Each report processed by the US DOE Technical Information Center is identified by a unique report number consisting of a code plus a sequential number. In most cases, the code identifies the originating installation. In some cases, it identifies a specific program or a type of publication. Listed in this publication are all codes that have been used by DOE in cataloging reports. This compilation consists of two parts. Part I is an alphabetical listing of report codes identified with the issuing installations that have used the codes. Part II is an alphabetical listing of installations identified with codes each has used

  18. Development of analytical code `ACCORD` for incore and plant dynamics of High Temperature Gas-cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Itakura, Hirofumi

    1996-11-01

    Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The `ACCORD` code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the `ACCORD` code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

  19. Development of analytical code 'ACCORD' for incore and plant dynamics of High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko; Itakura, Hirofumi.

    1996-11-01

    Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The 'ACCORD' code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the 'ACCORD' code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

  20. Remote-Handled Transuranic Content Codes

    International Nuclear Information System (INIS)

    2001-01-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document represents the development of a uniform content code system for RH-TRU waste to be transported in the 72-Bcask. It will be used to convert existing waste form numbers, content codes, and site-specific identification codes into a system that is uniform across the U.S. Department of Energy (DOE) sites.The existing waste codes at the sites can be grouped under uniform content codes without any lossof waste characterization information. The RH-TRUCON document provides an all-encompassing description for each content code and compiles this information for all DOE sites. Compliance with waste generation, processing, and certification procedures at the sites (outlined in this document foreach content code) ensures that prohibited waste forms are not present in the waste. The content code gives an overall description of the RH-TRU waste material in terms of processes and packaging, as well as the generation location. This helps to provide cradle-to-grave traceability of the waste material so that the various actions required to assess its qualification as payload for the 72-B cask can be performed. The content codes also impose restrictions and requirements on the manner in which a payload can be assembled. The RH-TRU Waste Authorized Methods for Payload Control (RH-TRAMPAC), Appendix 1.3.7 of the 72-B Cask Safety Analysis Report (SAR), describes the current governing procedures applicable for the qualification of waste as payload for the 72-B cask. The logic for this classification is presented in the 72-B Cask SAR. Together, these documents (RH-TRUCON, RH-TRAMPAC, and relevant sections of the 72-B Cask SAR) present the foundation and justification for classifying RH-TRU waste into content codes. Only content codes described in thisdocument can be considered for transport in the 72-B cask. Revisions to this document will be madeas additional waste qualifies for transport. Each content code uniquely

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

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

    International Nuclear Information System (INIS)

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

    2013-10-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

  3. SOURCE 1ST 2.0: development and beta testing

    International Nuclear Information System (INIS)

    Barber, D.H.; Iglesias, F.C.; Hoang, Y.; Dickson, L.W.; Dickson, R.S.; Richards, M.J.; Gibb, R.A.

    1999-01-01

    SOURCE 1ST 2.0 is the Industry Standard fission product release code that is being developed by Ontario Power Generation, New Brunswick Power, Hydro-Quebec, and Atomic Energy of Canada Ltd. This paper is a report on recent progress on requirement specification, code development, and module verification and validation activities. The theoretical basis for each model in the code is described in a module Software Theory Manual. The development of SOURCE IST 2.0 has required code design decisions about how to implement the software requirements. Development and module testing of the β1 release of SOURCE IST 2.0 (released in July 1999) have led to some interesting insights into fission product release modelling. The beta testing process has allowed code developers and analysts to refine the software requirements for the code. The need to verify physical reference data has guided some decisions on the code and data structure design. Examples of these design decisions are provided. Module testing, and verification and validation activities are discussed. These activities include code-targeted testing, stress testing, code inspection, comparison of code with requirements, and comparison of code results with independent algebraic, numerical, or semi-algebraic calculations. The list of isotopes to be modelled by SOURCE IST 2.0 provides an example of a subset of a reference data set. Isotopes are present on the list for a variety of reasons: personnel or public dose, equipment dose (for environmental qualification), fission rate and actinide modelling, or stable (or long-lived) targets for activation processes. To accommodate controlled changes to the isotope list, the isotope list and associated nuclear data are contained in a reference data file. The questions of multiple computing platforms, and of Year 2000 compliance have been addressed by programming rules for the code. By developing and testing modules on most of the different platforms on which the code is intended

  4. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  5. 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.

  6. 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

  7. A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration

    Directory of Open Access Journals (Sweden)

    Jensen Søren Holdt

    2005-01-01

    Full Text Available Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of audio signals. In this paper, we present a new perceptual model that predicts masked thresholds for sinusoidal distortions. The model relies on signal detection theory and incorporates more recent insights about spectral and temporal integration in auditory masking. As a consequence, the model is able to predict the distortion detectability. In fact, the distortion detectability defines a (perceptually relevant norm on the underlying signal space which is beneficial for optimisation algorithms such as rate-distortion optimisation or linear predictive coding. We evaluate the merits of the model by combining it with a sinusoidal extraction method and compare the results with those obtained with the ISO MPEG-1 Layer I-II recommended model. Listening tests show a clear preference for the new model. More specifically, the model presented here leads to a reduction of more than 20% in terms of number of sinusoids needed to represent signals at a given quality level.

  8. 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...

  9. GAM-HEAT -- a computer code to compute heat transfer in complex enclosures

    International Nuclear Information System (INIS)

    Cooper, R.E.; Taylor, J.R.; Kielpinski, A.L.; Steimke, J.L.

    1991-02-01

    The GAM-HEAT code was developed for heat transfer analyses associated with postulated Double Ended Guillotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re- radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices, and also accounts for convective, conductive, and advective heat exchanges. The code is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium. The GAM-HEAT code has been exercised extensively for computing transient temperatures in SRS reactors with specific charges and control components. Results from these computations have been used to establish the need for and to evaluate hardware modifications designed to mitigate results of postulated accident scenarios, and to assist in the specification of safe reactor operating power limits. The code utilizes temperature dependence on material properties. The efficiency of the code has been enhanced by the use of an iterative equation solver. Verification of the code to date consists of comparisons with parallel efforts at Los Alamos National Laboratory and with similar efforts at Westinghouse Science and Technology Center in Pittsburgh, PA, and benchmarked using problems with known analytical or iterated solutions. All comparisons and tests yield results that indicate the GAM-HEAT code performs as intended

  10. Great Expectations: Is there Evidence for Predictive Coding in Auditory Cortex?

    Science.gov (United States)

    Heilbron, Micha; Chait, Maria

    2017-08-04

    Predictive coding is possibly one of the most influential, comprehensive, and controversial theories of neural function. While proponents praise its explanatory potential, critics object that key tenets of the theory are untested or even untestable. The present article critically examines existing evidence for predictive coding in the auditory modality. Specifically, we identify five key assumptions of the theory and evaluate each in the light of animal, human and modeling studies of auditory pattern processing. For the first two assumptions - that neural responses are shaped by expectations and that these expectations are hierarchically organized - animal and human studies provide compelling evidence. The anticipatory, predictive nature of these expectations also enjoys empirical support, especially from studies on unexpected stimulus omission. However, for the existence of separate error and prediction neurons, a key assumption of the theory, evidence is lacking. More work exists on the proposed oscillatory signatures of predictive coding, and on the relation between attention and precision. However, results on these latter two assumptions are mixed or contradictory. Looking to the future, more collaboration between human and animal studies, aided by model-based analyses will be needed to test specific assumptions and implementations of predictive coding - and, as such, help determine whether this popular grand theory can fulfill its expectations. Copyright © 2017 The Author(s). Published by Elsevier Ltd.. All rights reserved.

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

    Science.gov (United States)

    Sikder, Somali; Ghosh, Shila

    2018-02-01

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

  12. Cost reducing code implementation strategies

    International Nuclear Information System (INIS)

    Kurtz, Randall L.; Griswold, Michael E.; Jones, Gary C.; Daley, Thomas J.

    1995-01-01

    Sargent and Lundy's Code consulting experience reveals a wide variety of approaches toward implementing the requirements of various nuclear Codes Standards. This paper will describe various Code implementation strategies which assure that Code requirements are fully met in a practical and cost-effective manner. Applications to be discussed includes the following: new construction; repair, replacement and modifications; assessments and life extensions. Lessons learned and illustrative examples will be included. Preferred strategies and specific recommendations will also be addressed. Sargent and Lundy appreciates the opportunity provided by the Korea Atomic Industrial Forum and Korean Nuclear Society to share our ideas and enhance global cooperation through the exchange of information and views on relevant topics

  13. Review and evaluation of technology, equipment, codes and standards for digitization of industrial radiographic film

    International Nuclear Information System (INIS)

    1992-05-01

    This reports contains a review and evaluation of the technology, equipment, and codes and standards related to the digitization of industrial radiographic film. The report presents recommendations and equipment-performance specifications that will allow the digitization of radiographic film from nuclear power plant components in order to produce faithful reproductions of flaw images of interest on the films. Justification for the specifications selected are provided. Performance demonstration tests for the digitization process are required and criteria for such tests is presented. Also several comments related to implementation of the technology are presented and discussed

  14. 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

  15. 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

  16. Gap Conductance model Validation in the TASS/SMR-S code using MARS code

    International Nuclear Information System (INIS)

    Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

    2010-01-01

    Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

  17. RFQ simulation code

    International Nuclear Information System (INIS)

    Lysenko, W.P.

    1984-04-01

    We have developed the RFQLIB simulation system to provide a means to systematically generate the new versions of radio-frequency quadrupole (RFQ) linac simulation codes that are required by the constantly changing needs of a research environment. This integrated system simplifies keeping track of the various versions of the simulation code and makes it practical to maintain complete and up-to-date documentation. In this scheme, there is a certain standard version of the simulation code that forms a library upon which new versions are built. To generate a new version of the simulation code, the routines to be modified or added are appended to a standard command file, which contains the commands to compile the new routines and link them to the routines in the library. The library itself is rarely changed. Whenever the library is modified, however, this modification is seen by all versions of the simulation code, which actually exist as different versions of the command file. All code is written according to the rules of structured programming. Modularity is enforced by not using COMMON statements, simplifying the relation of the data flow to a hierarchy diagram. Simulation results are similar to those of the PARMTEQ code, as expected, because of the similar physical model. Different capabilities, such as those for generating beams matched in detail to the structure, are available in the new code for help in testing new ideas in designing RFQ linacs

  18. 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

  19. Methodology for risk-based analysis of technical specifications

    International Nuclear Information System (INIS)

    Vesely, W.E.; Gaertner, J.P.; Wagner, D.P.

    1985-01-01

    Part of the effort by EPRI to apply probabilistic risk assessment methods and results to the solution of utility problems involves the investigation of methods for risk-based analysis of technical specifications. The culmination of this investigation is the SOCRATES computer code developed by Battelle's Columbus Laboratories to assist in the evaluation of technical specifications of nuclear power plants. The program is designed to use information found in PRAs to re-evaluate risk for changes in component allowed outage times (AOTs) and surveillance test intervals (STIs). The SOCRATES program is a unique and important tool for technical specification evaluations. The detailed component unavailability model allows a detailed analysis of AOT and STI contributions to risk. Explicit equations allow fast and inexpensive calculations. Because the code is designed to accept ranges of parameters and to save results of calculations that do not change during the analysis, sensitivity studies are efficiently performed and results are clearly displayed

  20. Improving coding accuracy in an academic practice.

    Science.gov (United States)

    Nguyen, Dana; O'Mara, Heather; Powell, Robert

    2017-01-01

    Practice management has become an increasingly important component of graduate medical education. This applies to every practice environment; private, academic, and military. One of the most critical aspects of practice management is documentation and coding for physician services, as they directly affect the financial success of any practice. Our quality improvement project aimed to implement a new and innovative method for teaching billing and coding in a longitudinal fashion in a family medicine residency. We hypothesized that implementation of a new teaching strategy would increase coding accuracy rates among residents and faculty. Design: single group, pretest-posttest. military family medicine residency clinic. Study populations: 7 faculty physicians and 18 resident physicians participated as learners in the project. Educational intervention: monthly structured coding learning sessions in the academic curriculum that involved learner-presented cases, small group case review, and large group discussion. overall coding accuracy (compliance) percentage and coding accuracy per year group for the subjects that were able to participate longitudinally. Statistical tests used: average coding accuracy for population; paired t test to assess improvement between 2 intervention periods, both aggregate and by year group. Overall coding accuracy rates remained stable over the course of time regardless of the modality of the educational intervention. A paired t test was conducted to compare coding accuracy rates at baseline (mean (M)=26.4%, SD=10%) to accuracy rates after all educational interventions were complete (M=26.8%, SD=12%); t24=-0.127, P=.90. Didactic teaching and small group discussion sessions did not improve overall coding accuracy in a residency practice. Future interventions could focus on educating providers at the individual level.

  1. Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

    Energy Technology Data Exchange (ETDEWEB)

    Protze, M. [Siemens-KWU, Erlangen (Germany)

    1995-12-31

    Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

  2. Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

    Energy Technology Data Exchange (ETDEWEB)

    Protze, M [Siemens-KWU, Erlangen (Germany)

    1996-12-31

    Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

  3. User's manual for a process model code

    International Nuclear Information System (INIS)

    Kern, E.A.; Martinez, D.P.

    1981-03-01

    The MODEL code has been developed for computer modeling of materials processing facilities associated with the nuclear fuel cycle. However, it can also be used in other modeling applications. This report provides sufficient information for a potential user to apply the code to specific process modeling problems. Several examples that demonstrate most of the capabilities of the code are provided

  4. Modular Modeling System (MMS) code: a versatile power plant analysis package

    International Nuclear Information System (INIS)

    Divakaruni, S.M.; Wong, F.K.L.

    1987-01-01

    The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry

  5. Validation of the TAC/BLOOST code (Contract research)

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

    2005-06-01

    Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30% (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test. (author)

  6. 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

  7. 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

  8. The PHREEQE Geochemical equilibrium code data base and calculations

    International Nuclear Information System (INIS)

    Andersoon, K.

    1987-01-01

    Compilation of a thermodynamic data base for actinides and fission products for use with PHREEQE has begun and a preliminary set of actinide data has been tested for the PHREEQE code in a version run on an IBM XT computer. The work until now has shown that the PHREEQE code mostly gives satisfying results for specification of actinides in natural water environment. For U and Np under oxidizing conditions, however, the code has difficulties to converge with pH and Eh conserved when a solubility limit is applied. For further calculations of actinide and fission product specification and solubility in a waste repository and in the surrounding geosphere, more data are needed. It is necessary to evaluate the influence of the large uncertainties of some data. A quality assurance and a check on the consistency of the data base is also needed. Further work with data bases should include: an extension to fission products, an extension to engineering materials, an extension to other ligands than hydroxide and carbonate, inclusion of more mineral phases, inclusion of enthalpy data, a control of primary references in order to decide if values from different compilations are taken from the same primary reference and contacts and discussions with other groups, working with actinide data bases, e.g. at the OECD/NEA and at the IAEA. (author)

  9. Through analysis of LOFT L2-3 by THYDE-P code

    International Nuclear Information System (INIS)

    Hirano, Masashi

    1981-10-01

    A through calculation of Experiment L2-3 of the Loss-of-Fluid Test (LOFT) Facility Power Ascension Series (Experiment Series L2) was performed with the THYDE-P code. The specific objectives of Experiment L2-3 were to determine the thermal-hydraulic behavior of the nuclear core and the thermal-mechanical response of the fuel rod cladding with a maximum linear heat generation rate of 39.4 kW/m. The THYDE-P code is a computer code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs) without a change in the methods and the models and is now under verification study and modification. The present calculation was performed by best estimate (BE) options as Sample Calculation Run 40, which is a portion of a series of THYDE-P sample calculations. The calculation was carried out from test initiation until complete submersion of the core volume with subcooled water, i.e. about 60 sec. The trend of the calculated cladding surface temperature was in good agreement with that of the experimental results. (author)

  10. User Effect on Code Application and Qualification Needs

    International Nuclear Information System (INIS)

    D'Auria, F.; Salah, A.B.

    2008-01-01

    Experience with some code assessment case studies and also additional ISPs have shown the dominant effect of the code user on the predicted system behavior. The general findings of the user effect investigations on some of the case studies indicate, specifically, that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under the general title of user effects. The specific characteristics of experimental facilities, i.e. limitations as far as code assessment is concerned; limitations of the used thermal-hydraulic codes to simulate certain system behavior or phenomena; limitations due to interpretation of experimental data by the code user, i.e. interpretation of experimental data base. On the basis of the discussions in this paper, the following conclusions and recommendations can be made: More dialogue appears to be necessary with the experimenters in the planning of code assessment calculations, e.g. ISPs.; User guidelines are not complete for the codes and the lack of sufficient and detailed user guidelines are observed with some of the case studies; More extensive user instruction and training, improved user guidelines, or quality assurance procedures may partially reduce some of the subjective user influence on the calculated results; The discrepancies between experimental data and code predictions are due both to the intrinsic code limit and to the so called user effects. There is a worthful need to quantify the percentage of disagreement due to the poor utilization of the code and due to the code itself. This need especially arises for the uncertainty evaluation studies (e.g. [18]) which do not take into account the mentioned user effects; A much focused investigation, based on the results of comparison calculations e.g. ISPs, analyzing the experimental data and the results of the specific code in order to evaluate the user effects and the related experimental aspects should be integral part of the

  11. Monte Carlo code for neutron radiography

    International Nuclear Information System (INIS)

    Milczarek, Jacek J.; Trzcinski, Andrzej; El-Ghany El Abd, Abd; Czachor, Andrzej

    2005-01-01

    The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms

  12. Monte Carlo code for neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Milczarek, Jacek J. [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)]. E-mail: jjmilcz@cyf.gov.pl; Trzcinski, Andrzej [Institute for Nuclear Studies, Swierk, 05-400 Otwock (Poland); El-Ghany El Abd, Abd [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland); Nuclear Research Center, PC 13759, Cairo (Egypt); Czachor, Andrzej [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)

    2005-04-21

    The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms.

  13. Static Verification for Code Contracts

    Science.gov (United States)

    Fähndrich, Manuel

    The Code Contracts project [3] at Microsoft Research enables programmers on the .NET platform to author specifications in existing languages such as C# and VisualBasic. To take advantage of these specifications, we provide tools for documentation generation, runtime contract checking, and static contract verification.

  14. Identification of coding and non-coding mutational hotspots in cancer genomes.

    Science.gov (United States)

    Piraino, Scott W; Furney, Simon J

    2017-01-05

    The identification of mutations that play a causal role in tumour development, so called "driver" mutations, is of critical importance for understanding how cancers form and how they might be treated. Several large cancer sequencing projects have identified genes that are recurrently mutated in cancer patients, suggesting a role in tumourigenesis. While the landscape of coding drivers has been extensively studied and many of the most prominent driver genes are well characterised, comparatively less is known about the role of mutations in the non-coding regions of the genome in cancer development. The continuing fall in genome sequencing costs has resulted in a concomitant increase in the number of cancer whole genome sequences being produced, facilitating systematic interrogation of both the coding and non-coding regions of cancer genomes. To examine the mutational landscapes of tumour genomes we have developed a novel method to identify mutational hotspots in tumour genomes using both mutational data and information on evolutionary conservation. We have applied our methodology to over 1300 whole cancer genomes and show that it identifies prominent coding and non-coding regions that are known or highly suspected to play a role in cancer. Importantly, we applied our method to the entire genome, rather than relying on predefined annotations (e.g. promoter regions) and we highlight recurrently mutated regions that may have resulted from increased exposure to mutational processes rather than selection, some of which have been identified previously as targets of selection. Finally, we implicate several pan-cancer and cancer-specific candidate non-coding regions, which could be involved in tumourigenesis. We have developed a framework to identify mutational hotspots in cancer genomes, which is applicable to the entire genome. This framework identifies known and novel coding and non-coding mutional hotspots and can be used to differentiate candidate driver regions from

  15. 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

  16. Genetic Code Analysis Toolkit: A novel tool to explore the coding properties of the genetic code and DNA sequences

    Science.gov (United States)

    Kraljić, K.; Strüngmann, L.; Fimmel, E.; Gumbel, M.

    2018-01-01

    The genetic code is degenerated and it is assumed that redundancy provides error detection and correction mechanisms in the translation process. However, the biological meaning of the code's structure is still under current research. This paper presents a Genetic Code Analysis Toolkit (GCAT) which provides workflows and algorithms for the analysis of the structure of nucleotide sequences. In particular, sets or sequences of codons can be transformed and tested for circularity, comma-freeness, dichotomic partitions and others. GCAT comes with a fertile editor custom-built to work with the genetic code and a batch mode for multi-sequence processing. With the ability to read FASTA files or load sequences from GenBank, the tool can be used for the mathematical and statistical analysis of existing sequence data. GCAT is Java-based and provides a plug-in concept for extensibility. Availability: Open source Homepage:http://www.gcat.bio/

  17. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  18. Calibration of the enigma code for Finnish reactor fuel with support from experimental irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Kelppe, S; Ranta-Puska, K [VTT Energy, Jyvaeskylae (Finland)

    1997-08-01

    Assessment by VTT of the ENIGMA fuel performance code, the original version by Nuclear Electric plc of the UK amended by a set of WWER specific materials correlations, is described. The given examples of results include analyses for BWR 9 x 9 fuel, BWR fuel irradiated in the reinstrumented test of an international Riso project, pre-characterized commercial WWER fuel irradiated in Loviisa reactor in Finland, and instrumented WWER test fuel irradiations in the MR reactor in Russia. The effects of power uncertainty and some model parameters are discussed. Considering the fact that the described cases all mean prototypic application of the code, the results are well encouraging. The importance of the accuracy in temperature calculations is emphasized. (author). 2 refs, 12 figs, 1 tab.

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

    International Nuclear Information System (INIS)

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

    1994-12-01

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

  20. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)