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Sample records for capability incorporating relap5

  1. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  2. PHISICS multi-group transport neutronic capabilities for RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

    2012-07-01

    PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

  3. Nuclear Hybrid Energy System Modeling: RELAP5 Dynamic Coupling Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Nolan Anderson; Haihua Zhao; Shannon Bragg-Sitton; George Mesina

    2012-09-01

    The nuclear hybrid energy systems (NHES) research team is currently developing a dynamic simulation of an integrated hybrid energy system. A detailed simulation of proposed NHES architectures will allow initial computational demonstration of a tightly coupled NHES to identify key reactor subsystem requirements, identify candidate reactor technologies for a hybrid system, and identify key challenges to operation of the coupled system. This work will provide a baseline for later coupling of design-specific reactor models through industry collaboration. The modeling capability addressed in this report focuses on the reactor subsystem simulation.

  4. RELAP5/MOD3 code coupling model

    International Nuclear Information System (INIS)

    A new capability has been incorporated into RELAP5/MOD3 that enables the coupling of RELAP5/MOD3 to other computer codes. The new capability has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 open-quotes styledclose quotes input and the Parallel Virtual Machine (PVM) software, which facilitates process management and distributed communication of multiprocess problems. RELAP5/MOD3 manages the input processing, communication instruction, process synchronization, and its own send and receive data processing. The flexible capability requires that an explicit coupling be established, which updates boundary conditions at discrete time intervals. Two test cases are presented that demonstrate the functionality, applicability, and issues involving use of this capability

  5. Incorporation of lithium lead eutectic as a working fluid in RELAP5 and preliminary safety assessment of LLCS

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K., E-mail: trivedi@iitk.ac.in [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India); Sandeep, K.T. [Institute for Plasma Research, Gandhinagar 382428 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India); Chaudhari, V.; Kumar, E.R. [Institute for Plasma Research, Gandhinagar 382428 (India); Munshi, P. [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India)

    2014-12-15

    Highlights: • The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in ITER. • It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. • RELAP steady state results closely match with the operating conditions of LLCS. • Results from transient calculations show that a maximum temperature of 875 K is attained 300 s after the loss of LLE flow. - Abstract: The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in International Thermonuclear Experimental reactor (ITER). It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. Lithium-lead eutectic (LLE) has been used as multiplier, breeder and coolant in TBM. Thermodynamic and transport properties of the LLE have been incorporated into the code. The main focus of this study is to check the heat transfer capability of LLE as coolant for TBM system for steady state and the considered anticipated operational occurrences (AOO's), namely, loss of heat source, loss of primary flow and loss of secondary flow. The six heat transfer correlation (reported for liquid metals in the literature) has been tested for steady state analysis of LLCS loop and results are roughly same for all of them. A good agreement has been observed between the operating conditions of LLCS with those of RELAP5 calculations. Results from transient calculations show that a maximum temperature of 875 K is attained during a 300 s loss of primary flow (LLE)

  6. RESTRUCTURING RELAP5-3D FOR NEXT GENERATION NUCLEAR PLANT ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; George L. Mesina; Joshua M. Hykes

    2006-06-01

    RELAP5-3D is used worldwide for analyzing nuclear reactors under both operational transients and postulated accident conditions. Development of the RELAP code series began in 1975 and since that time the code has been continuously improved, enhanced, verified and validated [1]. Since RELAP5-3D will continue to be the premier thermal hydraulics tool well into the future, it is necessary to modernize the code to accommodate the incorporation of additional capabilities to support the development of the next generation of nuclear reactors [2]. This paper discusses the reengineering of RELAP5-3D into structured code.

  7. RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa V. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: roseh@inrne.bas.bg; Stefanova, Antoaneta E. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin P. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg

    2005-08-15

    During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.

  8. RELAP5-3D User Problems

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan

    2002-09-01

    The Reactor Excursion and Leak Analysis Program with 3D capability1 (RELAP5-3D) is a reactor system analysis code that has been developed at the Idaho National Engineering and Environmental Laboratory (INEEL) for the U. S. Department of Energy (DOE). The 3D capability in RELAP5-3D includes 3D hydrodynamics2 and 3D neutron kinetics3,4. Assessment, verification, and validation of the 3D capability in RELAP5-3D is discussed in the literature5,6,7,8,9,10. Additional assessment, verification, and validation of the 3D capability of RELAP5-3D will be presented in other papers in this users seminar. As with any software, user problems occur. User problems usually fall into the categories of input processing failure, code execution failure, restart/renodalization failure, unphysical result, and installation. This presentation will discuss some of the more generic user problems that have been reported on RELAP5-3D as well as their resolution.

  9. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

    2012-10-01

    PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.

  10. Development Program of LOCA Licensing Calculation Capability with RELAP5-3D in Accordance with Appendix K of 10 CFR 50.46

    International Nuclear Information System (INIS)

    In light water reactors, particularly the pressurized water reactors, the severity of loss-of-coolant accidents (LOCAs) will limit how high the reactor power can extend. Although the best-estimate LOCA methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during LOCA, it will take many more resources to develop and to get final approval from the licensing authority. Instead, implementation of evaluation models required by Appendix K of the Code of Federal Regulations, Title 10, Part 50 (10 CFR 50), upon an advanced thermal-hydraulic platform can also gain significant margin on the PCT calculation. A program to modify RELAP5-3D in accordance with Appendix K of 10 CFR 50 was launched by the Institute of Nuclear Energy Research, Taiwan, and it consists of six sequential phases of work. The compliance of the current RELAP5-3D with Appendix K of 10 CFR 50 has been evaluated, and it was found that there are 11 areas where the code modifications are required to satisfy the requirements set forth in Appendix K of 10 CFR 50. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effect experiments and six sets of integral-effect experiments will be adopted. Through the assessments program, all the model changes will be verified

  11. Streamlining of the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Mesina, George L; Hykes, Joshua; Guillen, Donna Post

    2007-11-01

    RELAP5-3D is widely used by the nuclear community to simulate general thermal hydraulic systems and has proven to be so versatile that the spectrum of transient two-phase problems that can be analyzed has increased substantially over time. To accommodate the many new types of problems that are analyzed by RELAP5-3D, both the physics and numerical methods of the code have been continuously improved. In the area of computational methods and mathematical techniques, many upgrades and improvements have been made decrease code run time and increase solution accuracy. These include vectorization, parallelization, use of improved equation solvers for thermal hydraulics and neutron kinetics, and incorporation of improved library utilities. In the area of applied nuclear engineering, expanded capabilities include boron and level tracking models, radiation/conduction enclosure model, feedwater heater and compressor components, fluids and corresponding correlations for modeling Generation IV reactor designs, and coupling to computational fluid dynamics solvers. Ongoing and proposed future developments include improvements to the two-phase pump model, conversion to FORTRAN 90, and coupling to more computer programs. This paper summarizes the general improvements made to RELAP5-3D, with an emphasis on streamlining the code infrastructure for improved maintenance and development. With all these past, present and planned developments, it is necessary to modify the code infrastructure to incorporate modifications in a consistent and maintainable manner. Modifying a complex code such as RELAP5-3D to incorporate new models, upgrade numerics, and optimize existing code becomes more difficult as the code grows larger. The difficulty of this as well as the chance of introducing errors is significantly reduced when the code is structured. To streamline the code into a structured program, a commercial restructuring tool, FOR_STRUCT, was applied to the RELAP5-3D source files. The

  12. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.

  13. Architectural Advancements in RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L. Mesina

    2005-11-01

    As both the computer industry and field of nuclear science and engineering move forward, there is a need to improve the computing tools used in the nuclear industry to keep pace with these changes. By increasing the capability of the codes, the growing modeling needs of nuclear plant analysis will be met and advantage can be taken of more powerful computer languages and architecture. In the past eighteen months, improvements have been made to RELAP5-3D [1] for these reasons. These architectural advances include code restructuring, conversion to Fortran 90, high performance computing upgrades, and rewriting of the RELAP5 Graphical User Interface (RGUI) [2] and XMGR5 [3] in Java. These architectural changes will extend the lifetime of RELAP5-3D, reduce the costs for development and maintenance, and improve it speed and reliability.

  14. Presentation of RELAP5 results on the personal computer

    International Nuclear Information System (INIS)

    DrALF is a program for graphical presentation of RELAP5 results. Results may be displayed in two different forms, as graphs with different zoom capabilities and as drawings or nodalizations with different variables displayed on a background picture. (author)

  15. RELAP5/MOD2 overview and developmental assessment results from TMI-1 plant transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lin, J.C.; Tsai, C.C.; Ransom, V.H.; Johnsen, G.W.

    1984-01-01

    RELAP5/MOD2 is a new version of the RELAP5 thermal-hydraulic computer code containing improved modeling features that provide a generic capability for pressurized water reactor transient simulation. Objective of this paper is to provide code users with an overview of the code and to report developmental assessment results obtained from a Three Mile Island Unit One plant transient analysis. The assessment shows that the injection of highly subcooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2.

  16. BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code

    International Nuclear Information System (INIS)

    The TRACE code is still under development and it will have all capabilities of RELAP5. The purpose of the present study was therefore to assess the accuracy of the TRACE calculation of BETHSY 6.2TC test, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The overall results obtained with TRACE were similar to the results obtained by RELAP5/MOD3.3. The results show that the discrepancies were reasonable. (author)

  17. Recent SCDAP/RELAP5 code applications and improvements

    Energy Technology Data Exchange (ETDEWEB)

    Harvego, E.A.; Ghan, L.S.; Knudson, D.L.; Siefken, L.J. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

    1998-03-01

    This paper summarizes (1) a recent application of the severe accident analysis code SCDAP/RELAP5/MOD3.1, and (2) development and assessment activities associated with the release of SACDAP/RELAP5/MOD3.2. The Nuclear Regulatory Commission (NRC) has been evaluating the integrity of steam generator tubes during severe accidents. MOD3.1 has been used to support that evaluation. Studies indicate that the pressurizer surge line will fail before any steam generator tubes are damaged. Thus, core decay energy would be released as steam through the surge line and the tube wall would be spared from exposure to prolonged flow of high temperature steam. The latest code version, MOD3.2, contains several improvements to models that address both the early phase and late phase of a severe accident. The impact of these improvements to the overall code capabilities has been assessed. Results of the assessment are summarized in this paper.

  18. Data calculation program for RELAP 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Silvestre, Larissa J.B.; Sabundjian, Gaiane, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)

  19. SCDAP/RELAP5 independent peer review

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M.L. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering; Dhir, V.K. [Dhir, (V.K.) Santa Monica, CA (United States); Haste, T.J. [AEA Technology, Winfrith (United Kingdom); Heames, T.J. [Science Applications, Inc., Albuquerque, NM (United States); Jenks, R.P. [Los Alamos National Lab., NM (United States); Kelly, J.E. [Sandia National Labs., Albuquerque, NM (United States); Khatib-Rahbar, M. [Energy Research, Inc., Rockville, MD (United States); Viskanta, R. [Purdue Univ., Lafayette, IN (United States). Heat Transfer Lab.

    1993-01-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by recognized experts to determine overall technical adequacy, even though the code is still under development. For this purpose, an eight-member SCDAP/RELAP5 Peer Review Committee was organized, and the outcome of the review should help the NRC prioritize future code-development activity. Because the code is designed to be mechanistic, the Committee used a higher standard for technical adequacy than was employed in the peer review of the parametric MELCOR code. The Committee completed its review of the SCDAP/RELAP5 code, and the findings are documented in this report. Based on these findings, recommendations in five areas are provided: (1) phenomenological models, (2) code-design objectives, (3) code-targeted applications, (4) other findings, and (5) additional recommendations.

  20. SCDAP/RELAP5 independent peer review

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by recognized experts to determine overall technical adequacy, even though the code is still under development. For this purpose, an eight-member SCDAP/RELAP5 Peer Review Committee was organized, and the outcome of the review should help the NRC prioritize future code-development activity. Because the code is designed to be mechanistic, the Committee used a higher standard for technical adequacy than was employed in the peer review of the parametric MELCOR code. The Committee completed its review of the SCDAP/RELAP5 code, and the findings are documented in this report. Based on these findings, recommendations in five areas are provided: (1) phenomenological models, (2) code-design objectives, (3) code-targeted applications, (4) other findings, and (5) additional recommendations

  1. Validation of RELAP5 critical flow model

    International Nuclear Information System (INIS)

    The critical two-phase flow computerized simulation is made with the RELAP5 computer code. The methodology breaks the chocking process into one with either a two-phase inlet or a subcooled inlet. For two-phase flow the phases are assumed to be in thermal equilibrium. Thermal non-equilibrium is considered for subcooled upstream stagnation conditions. (authors); 6 refs., 3 figs

  2. Coupled RELAP5/GOTHIC model for IRIS SBLOCA analysis

    International Nuclear Information System (INIS)

    . However, the time that would be needed to develop new codes where all required models are properly incorporated, to build user experience, and to qualify the code would be prohibitively long and expensive. Therefore, the coupling of the thermal-hydraulic and containment codes can be an interesting approach and compromise between two modeling strategies mentioned above. Separate, existing computer codes can be coupled providing new capabilities without spending too much time in development and with possibility to use existing experience and perform code verification and validation only for the coupling portion of the new code. This coupling is usually performed as an extension of the classical calculation approach and it is localized at the physical points where communication between system and containment exists. For IRIS, it was decided to develop an explicit coupling of RELAP5/mod3.3, and one of the earlier versions of the GOTHIC code available at University of Zagreb, GOTHIC 3.4e; thus taking advantage of the rather large experience base in the use of the RELAP5 and GOTHIC codes as well as knowledge of their internal structure The primary goal was to explore applicability of coupled code to safety analyses of the new reactor systems where the primary system and containment closely interact. The chosen coupling strategy is simple and basic operation of constituent codes and corresponding input data are unaffected by the coupling process. This paper describes the coupled code as well as the development of the preliminary IRIS SBLOCA evaluation model and its use. Also, a discussion on the verification and validation of this methodology is provided.(author)

  3. Sub-channel analysis by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Alessandro Petruzzi; Anis Bousbia Salah [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2005-07-01

    Full text of publication follows: Recent progress in computer technology has increased the possibilities for code calculations in predicting realistically transient scenarios in nuclear power plants. Several attempts have been engaged in order to enlarge the domain for code applications, and to allow best estimate core simulation including interaction effects between neutronics and thermal-hydraulics. In this context, Relap5/Mod3.3 system thermalhydraulic code was used as a sub-channel code for the simulation of the low-pressure boil off experiment No 5002 of Neptun test facility. The experiment constitutes one of the separate effects test (SET) in the OECD/CSNI matrix for thermalhydraulic code validation related to phase separation and vertical flow 'with or without mixture level'. The drying out of the heated elements is expect to occur at very low coolant flow rates, low pressure (about 1.1 bar) and low power level (24.6 kW). The main aim of the activity discussed in the paper is to develop a 'nodalization technology' for accurately modeling the sub-channel grade void distribution problem and in the same way to assess the degree of success in using the Relap5 system code as a sub-channel code for the analysis of local quantities during transients in nuclear reactors. All thermal-hydraulic parameters, such as the collapsed liquid level, critical heat flux time occurrence and heaters surface temperature have been predicted with reasonable accuracy. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. More accurate results have been obtained considering the surface to surface radiation heat transfer model, as well as more cross flow nodes between the test section rods. The overall analysis confirms the possibility of using the Relap5/Mod3.3 system thermal-hydraulic code as sub-channel code to predict the evolution of relevant local quantities measured during 'relevant' experiments

  4. RELAP5-3D Restart and Backup Verification Testing

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2013-09-01

    Existing testing methodology for RELAP5-3D employs a set of test cases collected over two decades to test a variety of code features and run on a Linux or Windows platform. However, this set has numerous deficiencies in terms of code coverage, detail of comparison, running time, and testing fidelity of RELAP5-3D restart and backup capabilities. The test suite covers less than three quarters of the lines of code in the relap directory and just over half those in the environmental library. Even in terms of code features, many are not covered. Moreover, the test set runs many problems long past the point necessary to test the relevant features. It requires standard problems to run to completion. This is unnecessary for features can be tested in a short-running problem. For example, many trips and controls can be tested in the first few time steps, as can a number of fluid flow options. The testing system is also inaccurate. For the past decade, the diffem script has been the primary tool for checking that printouts from two different RELAP5-3D executables agree. This tool compares two output files to verify that all characters are the same except for those relating to date, time and a few other excluded items. The variable values printed on the output file are accurate to no more than eight decimal places. Therefore, calculations with errors in decimal places beyond those printed remain undetected. Finally, fidelity of restart is not tested except in the PVM sub-suite and backup is not specifically tested at all. When a restart is made from any midway point of the base-case transient, the restart must produce the same values. When a backup condition occurs, the code repeats advancements with the same time step. A perfect backup can be tested by forcing RELAP5 to perform a backup by falsely setting a backup condition flag at a user-specified-time. Comparison of the calculations of that run and those produced by the same input w/o the spurious condition should be

  5. Assessment of RELAP5/MOD3/K using semi scale test S-02-4

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sun Tack; Choi, Han Rim; Huh, Jae Yong; Lee, Nam Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report presents the results of the RELAP5/MOD3/K assessment utilizing a Semi scale large break loss-of-coolant experiment Test S-02-4. Blowdown heat transfer test S-02-4 is a 200 % double ended cold leg break experiment performed in Semi scale Mod-1 facility in 1975 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large break LOCA in a pressurized water reactor system. Through comparisons between data and best-estimate RELAP5 calculation, the capabilities of RELAP5 to calculate the large break loss-of-coolant accident (LOCA) were assessed. Emphasis was placed on the capability of the code to calculate break flow rates during system blowdown phase and the peak cladding temperature (PCT) behavior. 37 figs., 2 tabs., 7 refs. (Author) .new.

  6. Assessment of RELAP5/Mod3/KAERI using semiscale test S-06-3

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Choi, Han Rim; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report presents the results of the RELAP5/MOD3/KAERI assessment utilizing a semi scale large break loss-of-coolant experiment Test S-06-3. Test S-06-3 is a 200% double ended cold break experiment performed in semi scale Mod-1 facility in 1987 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large break LOCA in a pressurized were reactor system. Through comparisons between data and best-estimate RELAP5 calculation, the capabilities of RELAP5 to calculate the large break LOCA accident were assessed. Emphasis was placed on the capability of the code to calculate break flow rates during system blowdown phase, emergency core cooling system injection bypass during refill phase, quenching during reflood phase, and the peak cladding temperature behavior throughout the whole experiment. (Author) 12 refs., 38 figs., 2 tabs.

  7. Analysis of ROSA-III test RUN 704 by RELAP5/MOD0 code

    International Nuclear Information System (INIS)

    The ROSA-III test RUN 704 was analyzed for the assessment of RELAP5 code for BWR LOCA. RELAP5 is an advenced code developed to analyze thermal-hydraulic phenomena during LOCA and non-LOCA transients of LWR. It is based on a one-dimensional, nonhomogeneous, nonequilibrium two-phase flow model. The ROSA-III test RUN 704 is a standard BWR LOCA test, simulating a 200% double-ended break at the recirculation pump inlet pipe with all emergency core cooling systems activated. Large increase of core inlet flow due to lower plenum flashing and resulted rewetting of heater surface were calculated by RELAP5, indicating superior capability of RELAP5 two-phase flow model than RELAP4 phase separation model. Vapor and liquid counter-current flow was calculated at core inlet and core outlet. A small degree thermal nonequilibrium between vapor and liquid was calculated in the upper plenum after HPCS activation. However, core reflooding and quenching of heater surface were not calculated. There are still room for improvement in the interfacial drag and the heat transfer models of RELAP5/MOD0. (author)

  8. RELAP5-3D Code Includes ATHENA Features and Models

    International Nuclear Information System (INIS)

    Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF6, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)

  9. RELAP5-3D Code Includes Athena Features and Models

    Energy Technology Data Exchange (ETDEWEB)

    Richard A. Riemke; Cliff B. Davis; Richard R. Schultz

    2006-07-01

    Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, sf6, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5- 3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper.

  10. AUTOMATED, HIGHLY ACCURATE VERIFICATION OF RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    George L Mesina; David Aumiller; Francis Buschman

    2014-07-01

    Computer programs that analyze light water reactor safety solve complex systems of governing, closure and special process equations to model the underlying physics. In addition, these programs incorporate many other features and are quite large. RELAP5-3D[1] has over 300,000 lines of coding for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. Verification ensures that a program is built right by checking that it meets its design specifications. Recently, there has been an increased importance on the development of automated verification processes that compare coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions[2]. For the first time, the ability exists to ensure that the data transfer operations associated with timestep advancement/repeating and writing/reading a solution to a file have no unintended consequences. To ensure that the code performs as intended over its extensive list of applications, an automated and highly accurate verification method has been modified and applied to RELAP5-3D. Furthermore, mathematical analysis of the adequacy of the checks used in the comparisons is provided.

  11. Preliminary design report for SCDAP/RELAP5 lower core plate model

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

  12. Preliminary design report for SCDAP/RELAP5 lower core plate model

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Griffin, F.P. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

  13. RELAP5-3D multidimensional heat conduction enclosure model for RBMK reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Paik, S.

    1999-10-01

    A heat conduction enclosure model is conceived and implemented by RELAP5-3D between heat structures. The suggested model uses a lumped parameter model that is generally applicable to multidimensional calculational domain. This new model is applied to calculation of RBMK reactor core graphite blocks and is compared to the commercially available Fluid Dynamics Analysis Package (FIDAP) finite element code. Reasonably good agreement between the results of RELAP5-3D and FIDAP is obtained. The new heat conduction enclosure model gives RELAP5-3D a general multidimensional heat conduction capability. It also provides new routes for temperature cooloff of the RBMK graphite blocks from the ruptured channel to the surrounding ones. This ability to predict graphite temperature cooloff is very important during accidents or for transient simulation, especially concerning long-term coolability of the RBMK reactor core.

  14. Analysis of the reflood experiment by RELAP5/MOD2 code

    International Nuclear Information System (INIS)

    The analysis of the reflood experiment on the test rig Achilles has been performed. The analysis has been done by the RELAP5/MOD2 code after the results of the experiment had been released. The experiment has been analyze in several other laboratories around the world. Our results are comparable to other analyses and are in the range of RELAP5/MOD2 capabilities. Two analyses have been done: the core only and the complete system. Computed clad temperatures in the first case are higher than measured, in the second case they are somewhat lower. (author)

  15. RELAP5 analysis of two-phase decompression and rarefaction wave propagation under a temperature gradient

    International Nuclear Information System (INIS)

    The capability of RELAP5 to model single and two-phase acoustic waves is demonstrated with the use of fine temporal and spatial discretizations. Two cases were considered: a single phase air shock tube problem and pressure waves observed by Takeda and Toda in a two-phase decompression experiment in a pipe. Whereas the agreement for the single phase case is excellent, some discrepancies were observed in the two-phase case. However, RELAP5 produced markedly better results after adjusting the bubble size and the choked flow area. These results illustrate the need of a dynamic model for the interfacial area concentration (i.e., the bubble size). (author)

  16. Modification and validation of RELAP5/MOD3.2 for thermal-hydraulic accident analyses of HANARO

    International Nuclear Information System (INIS)

    Many aspects of RELAP5/MOD3.2 are modified by new features to properly simulate the HANARO characteristics such as the finned fuel elements and cooling by the plate type heat exchanger. Especially, RELAP5/MOD3.2 is modified a new heat transfer correlation package appropriate for the accident analysis of HANARO and developed plate type heat exchanger model. The validation of this newly developed code, RELAP5/HANARO, is carried out to assess its calculational capability to predict thermal-hydraulic behavior through the accident analysis in HANARO. The simulations performed are the single pin heat transfer and plate type heat exchanger experiments, which are then compared with experimental results and manufacturer's data, respectively. The simulation of natural convection experiment with the scaled bundle is also performed to evaluate the natural circulation cooling capability. The assessment results for the single pin heat transfer experiment and plate type heat exchanger model with RELAP5/HANARO showed conservatively good agreement with the experimental results. The natural circulation results calculated by RELAP5/HANARO are similar to the experimental data, even though minor discrepancies of the flow are identified. However, these differences are insignificant and conservatively acceptable thermal hydraulics. RELAP5/HANARO comprises the unique characteristics of HANARO and it is capable of simulating well the thermal-hydraulic behavior. In conclusion, it can be stated that this modified code provides a suitable analytical tool for the thermal-hydraulic accident analyses of HANARO

  17. Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations

    International Nuclear Information System (INIS)

    Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)

  18. RELAP5/MOD2 models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Dimenna, R.A.; Larson, J.R.; Johnson, R.W.; Larson, T.K.; Miller, C.S.; Streit, J.E.; Hanson, R.G.; Kiser, D.M.

    1988-08-01

    A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geothermal conditions; simplifying assumptions in implementation or application; and accuracy of implementation compared to documented descriptions of each of the models. An effort has been made to provide the reader with an understanding of what is in the code and why it is there and to provide enough information that an analyst can assess the impact of the correlation or model on the ability of the code to represent the physics of a reactor transient. Where assessment of the implemented versions of the models or correlations has been accomplished and published, the assessment results have been included.

  19. RELAP5/MOD2 models and correlations

    International Nuclear Information System (INIS)

    A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geothermal conditions; simplifying assumptions in implementation or application; and accuracy of implementation compared to documented descriptions of each of the models. An effort has been made to provide the reader with an understanding of what is in the code and why it is there and to provide enough information that an analyst can assess the impact of the correlation or model on the ability of the code to represent the physics of a reactor transient. Where assessment of the implemented versions of the models or correlations has been accomplished and published, the assessment results have been included

  20. Application of RELAP5 to a pipe blowdown experiment

    International Nuclear Information System (INIS)

    The application of the RELAP5 computer program to a pipe blowdown experiment is described in this paper. The basic hydrodynamic model, constitutive relations, and special process models included in RELAP5 are also briefly discussed. The results of this application confirm the effectiveness of using a choked flow model

  1. Independent review of SCDAP/RELAP5 natural circulation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, G.M.; Gross, R.J.; Martinez, M.J.; Rightley, G.S.

    1994-01-01

    A review and assessment of the uncertainties in the calculated response of reactor coolant system natural circulation using the SCDAP/RELAP5 computer code were completed. The SCDAP/RELAP5 calculation modeled a station blackout transient in the Surry nuclear power plant and concluded that primary system depressurization from natural circulation induced primary system failure is more likely than previously thought.

  2. Simulation of water hammer experiments using RELAP5 code

    International Nuclear Information System (INIS)

    The rapid closing or opening of a valve causes pressure transients in pipelines. The fast deceleration of the liquid results in high pressure surges upstream the valve, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increases. This phenomenon is called water hammer. The intensity of water hammer effects will depend upon the rate of change in the velocity or momentum. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the thermal-hydraulic system since, if the pressure induced exceeds the pressure range of a pipe given by the manufacturer, it can lead to the failure of the pipeline integrity. Due to its potential for damage of pipes, water hammer has been a subject of study since the middle of the nineteenth century. Many theoretical and experimental investigations were performed. The experimental investigation of the water hammer tests performed at Fraunhofer Institute for Environmental, Safety and Energy Technology (UMSICHT) [1] and Cold Water Hammer experiment performed by Forschungszentrum Rossendorf (CWHTF) [2] should be mentioned. The UMSICHT facility in Oberhausen was modified in order to simulate a piping system and associated supports that are typical for a nuclear power plant [3]. The Cold water hammer experiment is interesting and instructive because it covers a wide spectrum of particularities. One of them is sub-cooled water interaction with condensing steam at the closed end of the vertical pipe at room temperature and corresponding saturation pressure [4]. In the paper, the capabilities of RELAP5 code to correctly represent the water hammer phenomenon are presented. Paper presents the comparison of RELAP5 calculated and measured at UMSICHT and CWHTF test facilities pressure transient values after the fast closure (opening) of valves. The analyses of rarefaction wave travels inside the pipe and condensation of vapour bubbles in the liquid column

  3. MELCOR and SCDAP/RELAP5 code validation by simulation of TMI-2

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Chris [400 Central Drive, School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Liao, Yehong; Vierow, Karen

    2005-07-01

    Full text of publication follows: A comparison of the two severe accident codes, MELCOR and SCDAP/RELAP5, within the scope of thermal-hydraulic and core degradation models, has been previously performed by the authors for a hypothetical station blackout severe accident of a typical 4-loop PWR. This paper describes a validation of the codes. In the simulation of the TMI-2 severe accident, the data recorded during the accident and inferred from post-accident phenomena were used to investigate the soundness and compare the capabilities of MELCOR and SCDAP/RELAP5. With versatile control functions, best estimate thermal-hydraulic component models and detailed core models, MELCOR and SCDAP/RELAP5 produced similar predictions of the progression of TMI-2 accident, in simulating the actions of the plant control room operators, the system thermal-hydraulic response, the fuel damages, the core degradation and relocation. Input models and assumptions were modified to be as consistent yet true to the actual plant as possible. Due to different development approaches and some unique models, some minor discrepancies were observed between the predictions of MELCOR and SCDAP/RELAP5, which are within the uncertainties of the code numerical computation and the physics models. Some significant discrepancies in a few key areas were resolved either with a sophisticated method comparing with phenomena instead of raw code output information, or with a model modification based on actual plant data and experiment results. (authors)

  4. Advanced Presentation of BETHSY 6.2TC Test Results Calculated by RELAP5 and TRACE

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2012-01-01

    Full Text Available Today most software applications come with a graphical user interface, including U.S. Nuclear Regulatory Commission TRAC/RELAP Advanced Computational Engine (TRACE best-estimate reactor system code. The graphical user interface is called Symbolic Nuclear Analysis Package (SNAP. The purpose of the present study was to assess the TRACE computer code and to assess the SNAP capabilities for input deck preparation and advanced presentation of the results. BETHSY 6.2 TC test was selected, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The RELAP5 legacy input deck was converted to TRACE input deck using SNAP. The RELAP5 and TRACE comparison to experimental data showed that TRACE results are as good as or better than the RELAP5 calculated results. The developed animation masks were of great help in comparison of results and investigating the calculated physical phenomena and processes.

  5. Assessment of RELAP5/MOD2 and RELAP5/MOD1-EUR codes on the basis of LOBI-MOD2 test results

    International Nuclear Information System (INIS)

    The present report deals with an overview of the application of RELAP5/MOD2 and RELAP5/MOD1-EUR codes to tests performed in the LOBI/MOD2 facility. The work has been carried out in the frame of a contract between Dipartimento di Costruzioni Meccaniche e Nucleari (DCMN) of Pisa University and CEC. The Universities of Roma, Pisa, Bologna and Palermo and the Polytechnic of Torino performed the post-test analysis of the LOBI experiment under the supervision of DCMN. In the report the main outcomes from the analysis of the LOBI experiments are given with the attempt to identify deficiencies in the modelling capabilities of the used codes

  6. RELAP5 investigation on subchannel flow instability

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S.; Yang, B.W.; Liu, A.; Liu, X. [Xi' an Jiaotong Univ., Shaanxi (China). Science and Technology Center for Advanced Nuclear Fuel Research

    2016-07-15

    Two-phase flow instability is a vitally important area of study for a large number of industrial systems. Density Wave Oscillation (DWO) is the most common type of flow instability caused by the change in flow rate or power in boiling systems. The code RELAP5 is used to simulate single channel, 2 x 2 subchannels, and 3 x 3 subchannels with typical BWR subchannel geometry. The onset of flow instability determinating criterion and the results of simulations are utilized to create a stable boundary. The stable boundary of a single channel is compared with those from results of other researchers. Some conclusions are made as follows. 3 x 3 subchannels are more stable than single channel and 2 x 2 subchannels. Open subchannels possess a larger stable region than close channels. The heating model is analyzed determining that asymmetrical heating has negative effect on stability as compared to symmetric heating. With the analysis of transit time, period and subcooling number, there is a positive linear relationship between the subcooling number and oscillation period.

  7. RELAP5-3D Compressor Model

    Energy Technology Data Exchange (ETDEWEB)

    James E. Fisher; Cliff B. Davis; Walter L. Weaver

    2005-06-01

    A compressor model has been implemented in the RELAP5-3D© code. The model is similar to that of the existing pump model, and performs the same function on a gas as the pump performs on a single-phase or two-phase fluid. The compressor component consists of an inlet junction and a control volume, and optionally, an outlet junction. This feature permits cascading compressor components in series. The equations describing the physics of the compressor are derived from first principles. These equations are used to obtain the head, the torque, and the energy dissipation. Compressor performance is specified using a map, specific to the design of the machine, in terms of the ratio of outlet-to-inlet total (or stagnation) pressure and adiabatic efficiency as functions of rotational velocity and flow rate. The input quantities are specified in terms of dimensionless variables, which are corrected to stagnation density and stagnation sound speed. A small correction was formulated for the input of efficiency to account for the error introduced by assumption of constant density when integrating the momentum equation. Comparison of the results of steady-state operation of the compressor model to those of the MIT design calculation showed excellent agreement for both pressure ratio and power.

  8. Development of the unified version of COBRA/RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. J.; Ha, K. S.; Chung, B. D.; Lee, W. J.; Sim, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The COBRA/RELAP5 code, an integrated version of the COBRA-TF and RELAP5/MOD3 codes, has been developed for the realistic simulations of complicated, multi-dimensional, two-phase, thermal-hydraulic system transients in light water reactors. Recently, KAERI developed an unified version of the COBRA/RELAP5 code, which can run in serial mode on both workstations and personal computers. This paper provides the brief overview of the code integration scheme, the recent code modifications, the developmental assessments, and the future development plan. 13 refs., 5 figs., 2 tabs. (Author)

  9. Modeling of feed water check valves using RELAP5; Modellierung von Speisewasserrueckschlagventilen in RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Ben Said, Nader; Bregulla, Wolfgang; Kalk, Andreas [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2009-07-01

    Westinghouse Electric Germany GmbH has developed fluid dynamic models for medium-actuated armatures using the thermal hydraulic code RELAP5 in order to reach a more realistic description of the armature behavior including fluid-structure interactions in case of transient flow conditions in piping systems. The contribution is concerned with the modeling of damped check valves. The model allows the description of the behavior during opening and closure of a check armature. The calculated results show good agreement with the available measured data.

  10. Mathematic preprocessor for RELAP5 code using Microsoft Excel

    International Nuclear Information System (INIS)

    Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a toll for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-Excel software. (author)

  11. Transient simulation of ALWR passive safety systems using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Numerical simulation is presented of some passive safety systems currently incorporated in the design of the next generation advanced light water reactors (ALWRs). The performance and effectiveness of ex-core natural convection cooling and the concept of gravity driven water injection at high pressure are investigated using the RELAP5/MOD2 thermal-hydraulic code. The study identifies areas that should be investigated more fully in future experimental programs related to hypothetical large and small LOCA in ALWRs. (author)

  12. Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes

    Directory of Open Access Journals (Sweden)

    Viet-Anh Phung

    2015-01-01

    Full Text Available In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified.

  13. Improvement of RELAP5/MOD3.2.2 models for the development of CANDU auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Jung, B. D.; Lee, W. J.; Lim, H. S. [KAERI, Taejon (Korea, Republic of); Bang, Y. S.; Kim, M. W.; Lee, S. H. [KINS, Taejon (Korea, Republic of)

    1999-10-01

    Thermal-hydraulic models of the model improvements of NRC PWR auditing tool , i.e. RELAP5/MOD3, current auditing tool for LWR licensing, have been improved were attempted to develop a than auditingermal hydraulic auditing code for the CANDU. Major in order to identify the thermal hydraulic phenomena for the key CANDU events for key eventswere identified for reactor systems and components. Based on this, the modeling limitation of current RELAP5/MOD3 for CANDU applications were derived and the model improvement areas were identified. By improving these models in the code, RELAP5/MOD3/CANDU version has been developed. The new version was written in FORTRAN90 and its application capability has been verified through the simple verification calculations.

  14. Improvement of RELAP5/MOD3.2.2 models for the development of CANDU auditing code

    International Nuclear Information System (INIS)

    Thermal-hydraulic models of the model improvements of NRC PWR auditing tool , i.e. RELAP5/MOD3, current auditing tool for LWR licensing, have been improved were attempted to develop a than auditingermal hydraulic auditing code for the CANDU. Major in order to identify the thermal hydraulic phenomena for the key CANDU events for key eventswere identified for reactor systems and components. Based on this, the modeling limitation of current RELAP5/MOD3 for CANDU applications were derived and the model improvement areas were identified. By improving these models in the code, RELAP5/MOD3/CANDU version has been developed. The new version was written in FORTRAN90 and its application capability has been verified through the simple verification calculations

  15. Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

    Energy Technology Data Exchange (ETDEWEB)

    Papini, Davide, E-mail: davide.papini@mail.polimi.i [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Grgic, Davor [Department of Power Systems, FER, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Cammi, Antonio; Ricotti, Marco E. [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy)

    2011-04-15

    Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

  16. Development and assessment of the COBRA/RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Ha, Kwi Seok; Sim, Seok Ku

    1997-04-01

    The COBRA/RELAP5 code, a merged version of the COBRA-TF and RELAP5/MOD3.2 codes, has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This report provides the integration scheme of the two codes and the results of developmental assessments. These includes single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large-break loss-of-coolant experiment. From the single channel tests the integration scheme and its implementation were proven to be valid. Other simulation results showed good agreement with the experiments. The computational speed was also satisfactory. So it is confirmed that COBRA/RELAP5 can be a promising tool for analysis of complicated, multidimensional two-phase flow transients. The area of further improvements in the code integration are also identified. This report also serves as a user`s manual for the COBRA/RELAP5 code. (author). 6 tabs., 20 figs., 20 refs.

  17. The implementation of the CDC version of RELAP5/MOD1/019 on an IBM compatible computer system (AMDAHL 470/V8)

    International Nuclear Information System (INIS)

    RELAP5/MOD1 is an advanced one-dimensional best estimate system code, which is used for safety analysis studies of nuclear pressurized water reactor systems and related integral and separate effect test facilities. The program predicts the system response for large break, small break LOCA and special transients. To a large extent RELAP5/MOD1 is written in Fortran, only a small part of the program is coded in CDC assembler. RELAP5/MOD1 was developed on the CDC CYBER 176 at INEL*. The code development team made use of CDC system programs like the CDC UPDATE facility and incorporated in the program special purpose software packages. The report describes the problems which have been encountered when implementing the CDC version of RELAP5/MOD1 on an IBM compatible computer systems (AMDAHL 470/V8)

  18. SCDAP/RELAP5 code development and assessment

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Hohorst, J.K. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1996-03-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code`s calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities.

  19. Recent Improvements To The RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Richard A. Riemke; Paul D. Bayless; S. Michael Modro

    2006-06-01

    The RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) heat structures are allowed to be decoupled from hydrodynamic components, (2) built-in material properties for heat structures have been made consistent with those in MATPRO and the Nuclear Systems Materials Handbook (they are now documented in the RELAP5-3D manual, (3) Schrock's flow quality correlation is now used for a downward oriented junction from a horizontal volume for the stratification entrainment/pullthrough model.

  20. Study of turbine simulation model based on RELAP5

    International Nuclear Information System (INIS)

    The turbine model which can represent accurately non-isentropic process in the stage of turbine and system dynamic characteristics was developed and added into RELAP5 code, and the improvement of the turbine model of RELAP5 was implemented. The improved turbine model is based on the characteristics of steam flow and work transfer in the stage of turbine and considers adequately the impact of internal configuration parameters and oblique shock which is developed by non-equilibrium condensation of wet steam in turbine. Through building internal coupling interface and the modifying input processing subroutines, turbine model was developed as a part of RELAP5 hydro dynamic model. Taking the turbine of Qinshan 300 MW Nuclear Power Plant as an example, the simulation calculation and comparative analysis were performed for both stead and dynamic cases respectively by both the original and the modified turbine models in RELAP5 code. The results show that the modified turbine model can represent more accurately the dynamic operation characteristics of the turbine. (authors)

  1. RELAP5-3D Developer Guidelines and Programming Practices

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2014-03-01

    Our ultimate goal is to create and maintain RELAP5-3D as the best software tool available to analyze nuclear power plants. This begins with writing excellent programming and requires thorough testing. This document covers development of RELAP5-3D software, the behavior of the RELAP5-3D program that must be maintained, and code testing. RELAP5-3D must perform in a manner consistent with previous code versions with backward compatibility for the sake of the users. Thus file operations, code termination, input and output must remain consistent in form and content while adding appropriate new files, input and output as new features are developed. As computer hardware, operating systems, and other software change, RELAP5-3D must adapt and maintain performance. The code must be thoroughly tested to ensure that it continues to perform robustly on the supported platforms. The coding must be written in a consistent manner that makes the program easy to read to reduce the time and cost of development, maintenance and error resolution. The programming guidelines presented her are intended to institutionalize a consistent way of writing FORTRAN code for the RELAP5-3D computer program that will minimize errors and rework. A common format and organization of program units creates a unifying look and feel to the code. This in turn increases readability and reduces time required for maintenance, development and debugging. It also aids new programmers in reading and understanding the program. Therefore, when undertaking development of the RELAP5-3D computer program, the programmer must write computer code that follows these guidelines. This set of programming guidelines creates a framework of good programming practices, such as initialization, structured programming, and vector-friendly coding. It sets out formatting rules for lines of code, such as indentation, capitalization, spacing, etc. It creates limits on program units, such as subprograms, functions, and modules. It

  2. A generic semi-implicit coupling methodology for use in RELAP5-3D{copyright}

    Energy Technology Data Exchange (ETDEWEB)

    Aumiller, D.L.; Tomlinson, E.T.; Weaver, W.L.

    2000-09-01

    A generic semi-implicit coupling methodology has been developed and implemented in the RELAP5-3D{copyright} computer program. This methodology allows RELAP5-3D{copyright} to be used with other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The coupling methodology potentially allows different programs to be used to model different portions of the system. The programs are chosen based on their capability to model the phenomena that are important in the simulation in the various portions of the system being considered. The methodology was demonstrated using a test case in which the test geometry was divided into two parts each of which was solved as a RELAP5-3D{copyright} simulation. This test problem exercised all of the semi-implicit coupling features which were installed in RELAP5-3D0. The results of this verification test case show that the semi-implicit coupling methodology produces the same answer as the simulation of the test system as a single process.

  3. Using the RELAP5-3D advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. A demonstration calculation is presented. Finally, future tasks and plans are outlined. (author)

  4. Qualification of the code RELAP5/MOD2 with regard to pressurizer separate effect experiments

    International Nuclear Information System (INIS)

    As part of the RELAP5/MOD2 code assessment matrix, developed to qualify this simulation tool for analysis of pressurization transients in DWRs, two pressurizer separate effect experiments were analysed. The results allow some conclusions concerning simulation of pressurization transients in real nuclear plants. The geometry, thermal properties and heat losses of the pressurizer as well as the time lag constant of the instrumentation and the time step of the calculation were identified as the key parameters. Some conclusions were obtained concerning the code's capability to predict the thermal gradients, the heat transfer at the different interfaces, the condensation and evaporation rates, and their impact on pressure behaviour. (orig.)

  5. Qualification of the code RELAP5/MOD2 with regard to pressurizer separate effect experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1992-12-01

    As part of the RELAP5/MOD2 code assessment matrix, developed to qualify this simulation tool for analysis of pressurization transients in DWRs, two pressurizer separate effect experiments were analysed. The results allow some conclusions concerning simulation of pressurization transients in real nuclear plants. The geometry, thermal properties and heat losses of the pressurizer as well as the time lag constant of the instrumentation and the time step of the calculation were identified as the key parameters. Some conclusions were obtained concerning the code's capability to predict the thermal gradients, the heat transfer at the different interfaces, the condensation and evaporation rates, and their impact on pressure behaviour. (orig.).

  6. Modernization and restructuring of realistic thermal hydraulic system analysis code, RELAP5/MOD3.3.1.2

    International Nuclear Information System (INIS)

    feature is available for PC Windows users and provides simple Graphic User Interface (GUI) features. The productivity gains for both new, and experienced users from this userfriendly interface will be enormous, and the increased user productivity will pay back the developmental costs. RELAP5/MOD3.2.1.2 has been moderized and restructured in order to enhance the code portability, maintenance capability, readability, and flexibility. User convenience for PC Windows users has been realized by the on-line graphical processing through Windows programming. It should be noted that the code strcuture was fully domesticated, and future improvements could be easily carried out with the restructured version of RELAP5/MOD3.2.1.2

  7. RELAP5-3D Resolution of Known Restart/Backup Issues

    Energy Technology Data Exchange (ETDEWEB)

    Mesina, George L.; Anderson, Nolan A.

    2014-12-01

    The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequential verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.

  8. RL5SORT/RL5PLOT. A graphite package for the JRC-Ispra IBM version of RELAP5/MOD1

    International Nuclear Information System (INIS)

    The present report describes the programs RL5SORT and RL5PLOT, their implementation and ''how to use''. Both programs base on the IBM version of RELAP5 as developed at JRC-ISPRA. RL5SORT creates from the output file (restart plot file) of a RELAP5 calculation data files, which serve as input data base for the program RL5PLOT. RL5PLOT retrieves the previous stored data records (minor edit quantities of RELAP5), allows arithmetic operations with the retrieved data and enables a print or graphic output on the terminal screen of a TEKTRONIX graphic terminal. A set of commands, incorporated in the program RL5PLOT, facilitates the user's work. Program RL5SORT has been developed as a batch program, while RL5PLOT has been conceived for interactive working mode

  9. SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review

    Directory of Open Access Journals (Sweden)

    Andrea Achilli

    2012-01-01

    Full Text Available An Italian MSE R&D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

  10. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  11. Validation of RELAP5 with sensitivity analysis for uncertainty assessment for natural circulation two-phase flow instability

    International Nuclear Information System (INIS)

    The paper focuses on assessment of the capability of RELAP5 to predict natural circulation two-phase flow instability. The aim of this study is to identify needs for improvements of the code models. Experimental data from the low pressure CIRCUS facility was used for the code validation. The paper discusses the code validation procedure which combines separate and integral effect validation with the elements of the input errors propagation method for sensitivity analysis. The separate effect validation and 'transparent box' approach to the experimental system helps to identify main sources of experimental data uncertainty. The paper also discusses modifications provided in the new series of experiments to reduce the experimental uncertainty. Finally, the paper comes up with the conclusions about uncertainty in the RELAP5 prediction for different regimes of two-phase oscillatory flows in the CIRCUS facility and necessity for the models improvements. (author)

  12. RLP5SPL: a conversion program from RELAP5 output data to SPL format data

    International Nuclear Information System (INIS)

    A conversion program RLP5SPL has been developed, which converts RELAP5 output data to SPL format data. It has functions to extract plot informations from RELAP5 restart file and convert them to SPL data format. After conversion of RELAP5 output data, it is easy to perform unit conversion, comparisons with other calculational data and/or experimental data within figures, and plotting various type of figures including a bird-eye view of three dimensional surface. (author)

  13. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs

  14. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Johnson, E.C. (eds.); Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs.

  15. Peer review of RELAP5/MOD3 documentation

    International Nuclear Information System (INIS)

    A peer review was performed on a portion of the documentation of the RELAP5/MOD3 computer code. The review was performed in two phases. The first phase was a review of Vol. III, Developmental Assessment Problems, and Vol. IV, Models and Correlations. The reviewers for this phase were Dr. Peter Griffith, Dr. Yassin Hassan, Dr. Gerald S. Lellouche, Dr. Marino di Marzo and Mr. Mark Wendel. The reviewers recommended a number of improvements, including using a frozen version of the code for assessment guided by a validation plan, better discussion of discrepancies between the code and experimental data, and better justification for flow regime maps and extension of models beyond their data base. The second phase was a review of Vol. VI, Quality Assurance of Numerical Techniques in RELAP5/MOD3. The reviewers for the second phase were Mr. Mark Wendel and Dr. Paul T. Williams. Recommendations included correction of numerous grammatical and typographical errors and better justification for the use of Lax's Equivalence Theorem

  16. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Johnson, E.C. (eds.); Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code.

  17. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code

  18. ISP-46 analysis with RELAP5/SCDAPSIM computer code

    International Nuclear Information System (INIS)

    The thermal-hydraulic and severe accidents analysis code RELAP5/SCDAPSIM was used in the calculation of the Phebus FPT1 in-pile experiment. This experiment, carried out on 26 July 1996 in the Phebus facility, Cadarache, France, was chosen as the basis for the OECD International Standard Problem (ISP-46) exercise. Investigation of severe accidents phenomena like fuel degradation and hydrogen production was the objective of the ISP and of the presented analysis. The ISP was an open exercise, that is, all the relevant experimental results were available to the participants from the start. The FPT1 test bundle included 18 PWR fuel rods previously irradiated to a mean burnup of 23.4 GWd/tU, two instrumented fresh fuel rods and one silver-indium-cadmium control rod. The bundle was housed in an insulating shroud and introduced into the Phebus driver core which supplied the nuclear power. The fuel degradation phase of the test lasted about 5 hours during which the bundle was cooled by steam at pressure of about 2 bar with the mass flow rate varying between 0.5 g/s and 2.2 g/s, while the bundle nuclear power was being progressively increased from zero up to 36.5 kW. RELAP5/SCDAPSIM modelling of the Phebus facility and the main results, such as the temperature response of all rods and shroud, the oxidation and resulting hydrogen production, will be discussed and presented in this paper. The analysis of fuel rods degradation and problems related to SCDAPSIM underprediction of the amount of relocated fuel and cladding will also be covered. (author)

  19. Mathematic preprocessor for RELAP5 code using Microsoft Excel; Pre-processador matematico para o codigo RELAP5 utilizando o Microsoft Excel

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Patricia Andrea

    2006-07-01

    Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a toll for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-Excel software. (author)

  20. Comparative assessment of coupled RELAP5/PARCS and DYN3D/RELAP5 codes against the Kozloduy-6 pump trip test

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y.; Grundmann, U.; Kliem, S.; Rohde, U. [Institute of Safety Research, FZR, Dresden (Germany); Bousbia Salahn, A. [Pisa Univ., DIMNP (Italy)

    2005-07-01

    The modeling of complex transients in Nuclear Power Plants (NPP) remains a challenging topic for Best Estimate (BE) three-dimensional coupled code computational tools. Nowadays, this technique is extensively used since it allows decreasing conservatism in the calculation models by performing more realistic simulations based on a more precise consideration of multidimensional effects under complex transients in NPPs. This paper represents a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER-1000 pump trip test. The coupled RELAP5/3.3-PARCS/2.6 and DYN3D/3-RELAP5/3.3 code systems are used in simulations. The obtained results are assessed against experimental data and also through the code-to-code comparison. The DYN3D/RELAP5 computational model of VVER-1000 has been developed and adjusted for simulations with the parallel running scheme (PVM) of RELAP5/PARCS. Also, the macroscopic cross-section library used in the DYN3D/RELAP5 calculations has been adapted to meet the input requirements of PARCS. Prior to the test simulations, the RELAP5/PARCS model of the plant has been assessed in the stand-alone PARCS and RELAP5 test calculations. A reasonably good agreement between the experimental data and the calculated results is obtained. For the initial state, the observed discrepancies are mainly due to the absence of assembly discontinuity factor (ADF) correction and the evaluation of the Doppler feedback effect. During the transient, the deviations are mainly due to the combined effect of the measurement uncertainty in the control rod axial position and the estimation of the Doppler effect. (authors)

  1. Validation of RELAP5/MOD3.2 model on trip off one main coolant pump for VVER 440/V230

    Energy Technology Data Exchange (ETDEWEB)

    Groudev, Pavlin [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, Antoaneta [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg

    2006-06-15

    This paper presents the validation of RELAP5/MOD3.2 model of the VVER 440 for Nuclear Power Plant (NPP) in the analysis of the following transient: 'Trip off one MCP'. This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with measurement transient data received from Kozloduy NPP Unit no. 4. The baseline input deck for VVER440 was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events, and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparison between the RELAP5 calculations and the test data indicates a good agreement. This validation was possible through the participation of leading specialists from Kozloduy NPP and with the support of the Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

  2. High Flux Isotope Reactor system RELAP5 input model

    International Nuclear Information System (INIS)

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

  3. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  4. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and in this document, Volume II, to acquaint the user with the modeling base and thus aid in effective use of the code. 135 refs., 48 figs., 8 tabs

  5. Assessment of RELAP5-3D{copyright} using data from two-dimensional RPI flow tests

    Energy Technology Data Exchange (ETDEWEB)

    Davis, C.B.

    1998-07-01

    The capability of the RELAP5-3D{copyright} computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code`s logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved.

  6. PCRELAP5: data calculation program for RELAP 5 code; PCRELAP5: programa de calculo dos dados de entrada para o codigo RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Silvestre, Larissa Jacome Barros

    2016-07-01

    Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Calculo do RELAP5 - PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2. (author)

  7. 基于 RELAP5的汽轮机仿真模型研究%Study of Turbine Simulation Model Based on RELAP5

    Institute of Scientific and Technical Information of China (English)

    代守宝; 彭敏俊; 田兆斐; 姜昊

    2013-01-01

    建立了一个能准确反映级内部非等熵过程及动态运行特性的汽轮机模型,并将其加载到RELAP5程序中,完成RELAP5汽轮机模型的改进。改进的汽轮机模型是基于级内蒸汽的流动和做功特点,充分考虑了汽轮机结构参数以及汽轮机湿蒸汽流的非平衡两相凝结而形成的凝结冲波现象的影响。通过RELAP5程序内部耦合接口的建立和输入处理子程序的修改,实现了汽轮机模型的加载。以秦山一期300 MW核电厂汽轮机部件为对象,分别利用原RELAP5汽轮机模型和改进的汽轮机模型对其进行稳态和动态的仿真计算和比较分析。结果表明,改进的汽轮机模型能更准确地反映汽轮机动态运行特性。%The turbine model which can represent accurately non-isentropic process in the stage of turbine and system dynamic characteristics was developed and added into RELAP5 code ,and the improvement of the turbine model of RELAP5 was implemen-ted .The improved turbine model is based on the characteristics of steam flow and work transfer in the stage of turbine and considers adequately the impact of internal configura-tion parameters and oblique shock which is developed by non-equilibrium condensation of wet steam in turbine .Through building internal coupling interface and the modifying input processing subroutines ,turbine model was developed as a part of RELAP5 hydro-dynamic model . Taking the turbine of Qinshan 300 MW Nuclear Power Plant as an example ,the simulation calculation and comparative analysis were performed for both stead and dynamic cases respectively by both the original and the modified turbine mod-els in RELAP5 code .The results show that the modified turbine model can represent more accurately the dynamic operation characteristics of the turbine .

  8. International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-12-01

    Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes.

  9. RELAP5 Model Description and Validation for the BR2 Loss-of-Flow Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Koonen, E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-07-01

    This paper presents a description of the RELAP5 model, the calibration method used to obtain the minor loss coefficients from the available hydraulic data and the LOFA simulation results compared to the 1963 experimental tests for HEU fuel.

  10. Development of a VBA macro-based spreadsheet application for RELAP5 data post-processing

    Energy Technology Data Exchange (ETDEWEB)

    Belchior Junior, Antonio; Andrade, Delvonei A.; Sabundjian, Gaiane; Macedo, Luiz A.; Angelo, Gabriel; Torres, Walmir M.; Umbehaun, Pedro E.; Conti, Thadeu N., E-mail: abelchior@ipen.br, E-mail: delvonei@ipen.br, E-mail: gdjian@ipen.br, E-mail: lamacedo@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: tnconti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Bruel, Renata N. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)

    2011-07-01

    During the use of thermal-hydraulic codes such as RELAP5, large amount of data has to be managed in order to prepare its input data and also to analyze the produced results. This work presents a helpful tool developed to make it easier to handle the RELAP5 output data file. The XTRIP application is an electronic spreadsheet that contains some programmed macros that should be used for post-processing the RELAP5 output file. It can directly read the RELAP5 restart-plot binary output file and, through a user-friendly interface, transient results can be chosen and exported directly into an electronic worksheet. The XTRIP program can also do some data unit conversion as well as export these data to other programs such as Wingraf, Grapher and COBRA, etc. The main features of the developed Excel Visual Basic for Application macro as well as an example of use are presented and discussed. (author)

  11. Simulation of condensation in a closed, slightly inclined horizontal pipe with a modified RELAP5 code

    OpenAIRE

    Szijártó, Rita; Freixa Terradas, Jordi; Prasser, Horst-Michael

    2014-01-01

    The performance of the RELAP5 thermal-hydraulic system code was analyzed in predicting very fast transient condensation processes in horizontal pipes. The code significantly underpredicted the heat transfer from the primary to the secondary side in case of rapid wall condensation process in the so called Inverse Edwards Pipe Experiment, where the condensation pipe was immerged in a cool water pool, and hot steam injection was performed into a pipe, which was closed on one side. The RELAP5 con...

  12. RELAP5 assessment to IAEA HTR-10 benchmark problem-I

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. S.; Jung, B. J. [Cheju National University, Cheju (Korea, Republic of); Lee, W. J.; Jung, B. D. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    In order to evaluate the capability of RELAP5 code, a system thermal-hydraulic safety analysis code for water reactors, for the analysis of HTGR (High Temperature Gas Cooled Reactor) RCCS, an IAEA Benchmark Problem-I for the HTR-10 was assessed. Assessment results were compared with the results of THERMIX code, a thermal-hydraulic analysis code for HTGR. The calculated results showed good agreement with those by the THERMIX code with a maximum deviation around 4.5%. Deviation was evaluated to originate from the simplification of complicated geometry and from the modeling capability of heat transfer characteristics in the HTGR components such as water cooler and air cooler. Especially, it was found that the radiation heat transfer in the reactor cavity played an important role in the after heat removal by the RCCS. Thus, it is concluded that it is necessary to evaluate and improve relevant models both for the convection and radiation heat transfer in order to enhance the code analysis capability to the HTGR.

  13. Assessment of a RELAP5 model for the IPR-R1 Triga research reactor

    International Nuclear Information System (INIS)

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this way, as a contribution to the assessment of RELAP5/3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed by a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open-pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data and also calculation data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code were considered in the process of the model validation. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual reactor behavior in good agreement with the available data. (author)

  14. Development and validation of coupled PARCS/RELAP5 model for Forsmark NPP at uprated power

    International Nuclear Information System (INIS)

    This paper gives an account of the development and validation of an up-to-date coupled neutronic/thermal-hydraulic model for the Swedish Forsmark boiling water reactor. The model will be used for analyses of the consequences of the planned power uprate from 2928 MWth to 3253 MWth. At first, the development of the PARCS and RELAP5 models are presented. On the neutronic side, cross-sections data was generated, allowing feeding PARCS with realistic data. This step was performed by converting the library data file from the power plant using the in-house cross-section interface code. The dependence of the material properties on history effects, burnup, and instantaneous conditions was accounted for, and the full heterogeneity of the core was thus taken into account. Each of the 676 fuel assemblies was modeled individually, while the 161 control rods were grouped into 6 different types. On the thermal-hydraulic side, the model consists of a model for the feedwater system, a model for the reactor vessel that include a model for the core channels, and a model for each of the four steam lines. The fuel assemblies were modeled as twelve flow channels in the core region. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. The validation efforts were focusing on benchmarking the code capabilities against measured plant data, both under steady-state and transient conditions. The PARCS standalone model was validated against traversing in-core probe (TIP) measurements, taken at different burnup level with operating power varies from 108% (nominal level) to 120% (uprated level). The coupled PARCS/RELAP5 model was validated against an operational transient. For this validation task, the transient chosen was a turbine trip test, which was performed on May 6, 2013. Comparisons between calculated and measured parameters demonstrate that the coupled model was able to correctly represent the

  15. RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom

    2012-06-01

    The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.

  16. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  17. A RELAP5 study to identify flow regime in natural circulation phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Torres, Walmir M.; Macedo, Luiz A.; Mesquita, Roberto N.; Andrade, Delvonei A.; Umbehaun, Pedro E.; Conti, Thadeu N.; Masotti, Paulo H.F.; Belchior Junior, Antonio; Angelo, Gabriel, E-mail: gdjian@ipen.b, E-mail: umbehaun@ipen.b, E-mail: wmtorres@ipen.b, E-mail: tnconti@ipen.b, E-mail: rnavarro@ipen.b, E-mail: lamacedo@ipen.b, E-mail: pmasotti@ipen.b, E-mail: abelchior@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    There has been a crescent interest in the scientific community in the study of natural circulation phenomenon. New generation of compact nuclear reactors uses the natural circulation of the fluid as a system of cooling and of residual heat removal in case of accident or shutdown. The objective of this paper is to compare the flow patterns of experimental data and numerical simulation for the natural circulation phenomenon in two-phase flow regime. An experimental circuit built with glass tubes is used for the experiments. Thus, it allows the thermal hydraulic phenomena visualization. There is an electric heater as the heat source, a heat exchanger as the heat sink and an expansion tank to accommodate fluid density excursions. The circuit instrumentation consists of thermocouples and pressure meters to better keep track of the flow and heat transfer phenomena. Data acquisition is performed through a computer interface developed with LABVIEW. The characteristic of the regime is identified using photography techniques. Numerical modeling and simulation is done with the thermal hydraulic code RELAP5, which is widely used for this purpose. This numerical simulation is capable to reproduce some of the flow regimes which are present in the circuit for the natural circulation phenomenon. Comparison between experimental and numerical simulation is presented in this work. (author)

  18. The assessment of RELAP5/MOD2 based on pressurizer transient experiments

    International Nuclear Information System (INIS)

    Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

  19. Parametric study of different perturbations on Ringhals stability benchmark with RELAP5/PARCS

    International Nuclear Information System (INIS)

    The analysis of power instabilities were tackled many years ago developing new methodologies to model this phenomena. The mechanisms underlying the causes of the power oscillations in BWR are still under study, but its consequences are well known. The simulation of the instabilities using best-estimate codes is the aim of this work. Three dimensional time domain BWR stability analysis has been performed in Ringhals 1 NPP, using the coupled code RELAP5-MOD3.3/PARCS v2.7. In the simulation of instabilities, it is necessary to introduce some perturbations that make the power oscillate. In this work, the instabilities are induced by means of density perturbations using a new capability introduced in the neutronic code. The applied perturbation is based on the Lambda modes and their amplitudes. This new option permits the user to perturb the moderator density in each node at each time step. Using different amplitudes for the perturbation signal the user is able to perform a complete stability analysis studying the resulting power oscillations. (author)

  20. RELAP5/MOD3 code manual. Volume 4, Models and correlations

    International Nuclear Information System (INIS)

    The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations

  1. Implementation of DOWTHERM A Properties into RELAP5-3D/ATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Richard L. Moore

    2010-04-01

    DOWTHERM A oil is being considered for use as a heat transfer fluid in experiments to help in the design of heat transfer components for the Next Generation Nuclear Plant (NGNP). In conjection with the experiments RELAP5-3D/ATHENA will be used to help design and analyzed the data generated by the experiments. Inorder to use RELAP5-3D the thermophysical properties of DOWTHERM A were implemented into the fluids package of the RELAP5-3D/ATHENA computer propgram. DOWTHERM A properties were implemented in RELAP5-3D/ATHENA using thermophysical property data obtain from a Dow Chemical Company brochure. The data were curve fit and the polynomial equations developed for each required property were input into a fluid property generator. The generated data was then compared to the orginal DOWTHERM A data to verify that the fluid property data generated by the RELAP5-3D/ATHENA code was representitive of the original input data to the generator.

  2. RELAP5/MOD3 code manual. Volume 4, Models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations.

  3. Assessment and improvement of condensation models in RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

  4. RELAP5 simulation for one and two-phase natural circulation phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de; Umbehaun, Pedro Ernesto; Torres, Walmir Maximo; Castro, Alfredo Jose Alvim de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: gdjian@ipen.br; delvonei@ig.com.br; umbehaun@ipen.br; wmtorres@ipen.br; Braz Filho, Francisco A.; Borges, Eduardo Madeira [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mails: eduardo@ieav.cta.br; fbraz@ieav.cta.br; Belchior Junior, Antonio; Rocha, Ricardo Takeshi Vieira da [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)]. E-mails: belchior@bol.com.br; rtvrocha@uol.com.br; Damy, Osvaldo Luiz Almeida; Torres, Eduardo [Universidade de Sao Paulo (USP), SP (Brazil). Escola Politecnica]. E-mails: osvaldo.damy@poli.usp.br; etorres@pac.ind.br

    2007-07-01

    The objective of this paper is to study the natural circulation phenomenon in one and two-phase regime. There has been a crescent interest in the scientific community in the study of the natural circulation. New generation of compact nuclear reactors uses the natural circulation for residual heat removal in case of accident or shutdown. For this study, the modeling and the simulation of the experimental circuit is performed with the RELAP5 code. The experimental circuit is mounted in the Chemical Engineering Department of the University of Sao Paulo. It is presented in this work the theoretical/experimental comparison for one and two-phase flow. These results will be stored in a database to validate RELAP5 calculations. This work was also used to training some users of RELAP5 from IEAv. (author)

  5. RELAP5 simulation for one and two-phase natural circulation phenomenon

    International Nuclear Information System (INIS)

    The objective of this paper is to study the natural circulation phenomenon in one and two-phase regime. There has been a crescent interest in the scientific community in the study of the natural circulation. New generation of compact nuclear reactors uses the natural circulation for residual heat removal in case of accident or shutdown. For this study, the modeling and the simulation of the experimental circuit is performed with the RELAP5 code. The experimental circuit is mounted in the Chemical Engineering Department of the University of Sao Paulo. It is presented in this work the theoretical/experimental comparison for one and two-phase flow. These results will be stored in a database to validate RELAP5 calculations. This work was also used to training some users of RELAP5 from IEAv. (author)

  6. RHF RELAP5 model and preliminary loss-of-offsite-power simulation results for LEU conversion

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Bergeron, A. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Dionne, B. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Thomas, F. [Institut Laue-Langevin (ILL), Grenoble (Switzerland). RHF Reactor Dept.

    2014-08-01

    The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

  7. RHF RELAP5 Model and Preliminary Loss-Of-Offsite-Power Simulation Results for LEU Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Bergeron, A. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Dionne, B. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Thomas, F. [Institut Laue-Langevin (ILL), Grenoble (Switzerland). RHF Reactor Dept.

    2014-08-01

    The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

  8. Assessment of RELAP5/MOD3.1 for gravity-driven injection experiment in the core makeup tank of the CARR Passive Reactor (CP-1300)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.I.; No, H.C. [Korea Advanced Inst. of Science and Technology, Yusung, Taejon (Korea, Republic of). Nuclear Engineering Dept.; Bang, Y.S.; Kim, H.J. [Korea Inst. of Nuclear Safety, Yusung Taejon (Korea, Republic of). Advanced Reactor Dept.

    1996-10-01

    The objective of the present work is to improve the analysis capability of RELAP5/MOD3.1 on the direct contact condensation in the core makeup tank (CMT) of passive high-pressure injection system (PHPIS) in the CARR Passive Reactor (CP-1300). The gravity-driven injection experiment is conducted by using a small scale test facility to identify the parameters having significant effects on the gravity-driven injection and the major condensation modes. It turns out that the larger the water subcooling is, the more initiation of injection is delayed, and the sparger and the natural circulation of the hot water from the steam generator accelerate the gravity-driven injection. The condensation modes are divided into three modes: sonic jet, subsonic jet, and steam cavity. RELAP5/MOD3.1 is chosen to evaluate the cod predictability on the direct contact condensation in the CMT. It is found that the predictions of MOD3.1 are in better agreement with the experimental data than those of MOD3.0. From the nodalization study of the test section, the 1-node model shows better agreement with the experimental data than the multi-node models. RELAP5/MOD3.1 identifies the flow regime of the test section as vertical stratification. However, the flow regime observed in the experiment is the subsonic jet with the bubble having the vertical cone shape. To accurately predict the direct contact condensation in the CMT with RELAP5/MOD3.1, it is essential that a new set of the interfacial heat transfer coefficients and a new flow regime map for direct contact condensation in the CMT be developed.

  9. RELAP5-3D{sup C} three dimensional neutron kinetics coupled thermal-hydraulics analyses of the Atucha-2 PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, C.; D' Auria, F. [GRNSPG / University of Pisa, Pisa (Italy); Mazzantini, O. [NA-SA, Buenos Aires (Argentina); Ivanov, K.N. [RDFMG / The Pennsylvania State University, State College, PA (United States)

    2008-07-01

    In the framework of the Agreement 'NA-SA - University of Pisa', a detailed RELAP5-3D{sup C} model of the under construction Atucha-2 PHWR was developed. The aim of this activity was to have a state-of-the-art tool for performing realistic safety analyses for the licensing of Atucha-2. In this paper, the main steps for the development and the qualification of the RELAP5-3D{sup C} model as well as some sample applications for the safety analyses (0.1 LOCA, Fuel Channel blockage) are reported. After the definition of a relevant core status with fuel at the burnup equilibrium, a set of neutron cross section libraries were calculated by the lattice physics code HELIOS. These data were used for the setting up of a RELAP5-3D{sup C} Neutron Kinetic (NK) model of the core that was then coupled with a RELAP5-3D{sup C} Thermal-Hydraulic (TH) model of the whole plant. The TH model is based on a 280 Fuel Channels (FCs) and a 3D moderator tank nodalization. The 3D NK model is representing all the FCs, the reflectors and the oblique Control Rods (CR). The boundary conditions for the reconstruction of the boron clouds, injected in the moderator tank by the shut-down emergency system, were derived by previously executed Computational Fluid-dynamics (CFD) analyses by the CFX code. Finally, applications of the developed 3D NK-TH model to sample safety analyses demonstrated its state-of-the-art capabilities and its high level of realism. (authors)

  10. Design report on SCDAP/RELAP5 model improvements - debris bed and molten pool behavior

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Rempe, J.L.; Chavez, S.A.

    1994-11-01

    the SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and in combination with VICTORIA, fission product release and transport during severe accidents. Improvements for existing debris bed and molten pool models in the SCDAP/RELAP5/MOD3.1 code are described in this report. Model improvements to address (a) debris bed formation, heating, and melting; (b) molten pool formation and growth; and (c) molten pool crust failure are discussed. Relevant data, existing models, proposed modeling changes, and the anticipated impact of the changes are discussed. Recommendations for the assessment of improved models are provided.

  11. Assessment and improvement of condensation model in RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Rho, Hui Cheon; Choi, Kee Yong; Park, Hyeon Sik; Kim, Sang Jae [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Lee, Sang Il [Korea Power Engineering Co., Inc., Seoul (Korea, Republic of)

    1997-07-15

    The objective of this research is to remove the uncertainty of the condensation model through the assessment and improvement of the various heat transfer correlations used in the RELAP5/MOD3 code. The condensation model of the standard RELAP5/MOD3 code is systematically arranged and analyzed. A condensation heat transfer database is constructed from the previous experimental data on various condensation phenomena. Based on the constructed database, the condensation models in the code are assessed and improved. An experiment on the reflux condensation in a tube of steam generator in the presence of noncondensable gases is planned to acquire the experimental data.

  12. Thermal hydraulic analysis of the multipurpose research reactor RMB using a RELAP5 model

    International Nuclear Information System (INIS)

    The Multipurpose Brazilian Reactor (RMB) will be an open pool multipurpose research reactor using low enriched uranium fuel (LEU). This paper presents the RMB nodalization and the first thermal hydraulic results of steady state calculations using the RELAP5-MOD3.3 code. Several current investigations have shown that RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions in spite of such code was initially projected to studies of commercial nuclear power plants. (author)

  13. A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, M. [Swiss Federal Institute of Technology, Zurich (Switzerland); Analytis, G.T.; Aksan, S.N. [Paul Scherrer Institute, Villigen (Switzerland)

    1995-09-01

    The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for the predetermined position of the quench front. Both RELAP5/MOD2.5 and RELAP5/MOD3 substantially underpredicted the exit vapour temperature. The mass flow rate and quality, however, were correct and the heat input to the vapour was larger than the actual one. As the vapour superheat at the tube exit depends on the balance between the heat input from the wall and the heat exchange with the droplets, the discrepancy between the calculated and the measured exit vapour temperature suggested that the inability of both codes to predict the vapour superheat in the dispersed flow region is due to the overprediction of the interfacial heat transfer rate.

  14. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    International Nuclear Information System (INIS)

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and non-porous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of non-porous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  15. SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

    International Nuclear Information System (INIS)

    Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head during a severe accident in a Light Water Reactor. Currently, the COUPLE model has no capability to model the movement of material that melts within a matrix of porous material. The COUPLE model also does not have the capability to model the movement of liquefied core plate material that slumps onto a porous debris bed in the lower head. In order to advance beyond the assumption the liquefied material always remains stationary, designs are developed for calculations of the movement of liquefied material through the interstices in a matrix of porous material. Correlations are identified for calculating the permeability of the porous debris and for calculating the rate of flow of liquefied material through the interstices in the debris bed. Correlations are also identified for calculating the relocation of solid debris that has a large amount of cavities due to the flowing away of melted material. Equations are defined for calculating the effect on the temperature distribution in the debris bed of heat transported by moving material and for changes in effective thermal conductivity and heat capacity due to the movement of material. The implementation of these models is expected to improve the calculation of the material distribution and temperature distribution of debris in the lower head for cases in which the debris is porous and liquefied material is present within the porous debris

  16. Extending and Verification of RELAP5 Code for Liquid Fueled Molten Salt Reactor%RELAP5应用于液态燃料熔盐堆的扩展及验证

    Institute of Scientific and Technical Information of China (English)

    施承斌; 程懋松; 刘桂民

    2016-01-01

    为将RELAP5程序应用于液态燃料熔盐堆的建模分析,需要对RELAP5的模型进行扩展.基于RELAP5原有的点堆模型和热工水力模型,新增了液态燃料熔盐堆点堆模型和相应的带有内热源的热工水力模型,并通过熔盐实验堆(MSRE)实验数据进行验证.结果表明,扩展后的RELAP5程序能够适用于液态燃料熔盐堆系统建模分析.%In order to analyze the liquid fueled molten salt reactor using RELAP5 code,models in RELAP5 code need to be extended.This paper attempts to add new point kinetic model of liquid fueled reactor and thermo-hydraulics model with internal heat source based on the original RELAP5 models,then the code is verified using MSRE experimental datum.The results indicate that the extended RELAP5 code can be applied to model and analyze the liquid fueled molten salt reactor.

  17. Sub-channel analysis by RELAP5 system code of boil-off experiment (Test 5002) with NEPTUN facility

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A. [Pennsylvania State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, Pennsylvania (United States)]. E-mail: axp46@psu.edu; Bousbia Salah, A.; D' Auria, F. [Univ. of Pisa, Dipartimento di Ingegneria Meccanica, Nucleare d della Produzione, Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; f.dauria@ing.unipi.it

    2004-07-01

    This paper presents the results of RELAP5/Mod3.2 system thermalhydraulic code using the sub-channel analysis approach in predicting the NEPTUN separate effect boil off experiments. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the NEPTUN low pressure test N{sup o}5002 has been considered. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory and demonstrate, as well, the reasonable success of the 'sub-channel analysis' approach adopted in the present context for a system thermalhydraulic code. (author)

  18. Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code

    Science.gov (United States)

    De Rosa, F.; Lombardo, C.; Mascari, F.; Polidori, M.; Chiovaro, P.; D'Amico, S.; Moscato, I.; Vella, G.

    2014-11-01

    The present paper deals with the investigation of the evolution and consequences of a Station Black-Out (SBO) initiating event transient in the SPES3 facility [1]. This facility is an integral simulator of a small modular reactor being built at the SIET laboratories, in the framework of the R&D program on nuclear fission funded by the Italian Ministry of Economic Development and led by ENEA. The SBO transient will be simulated by using the RELAP5 and TRACE nodalizations of the SPES3 facility. Moreover, the analysis will contribute to study the differences on the code predictions considering the different modelling approach with one and/or three-dimensional components and to compare the capability of these codes to describe the SPES3 facility behaviour.

  19. RELAP5 code validation using a medium-size break LOCA experiment at the PMK-2 test facility

    International Nuclear Information System (INIS)

    For the analyses of loss of coolant accidents (LOCA) the thermohydraulic computer code capabilities for eastern-type reactors like VVER-440 must be validated by pre- and post test calculations of suitable experiments. Such experiments are performed on PMK-2 integral-type test facility in KFKI Atomic Energy Research Institute, Budapest, which is a volume-scaled model of the primary and secondary system of the Paks Nuclear Power Plant. One of these experiments is the pressuriser surge line break which correspond to a 22% leak. The most important phenomena of the experiment are the behavior of hot leg loop seal and the core dry-out with refill-reflood. Posttest calculations were performed by use of the code version RELAP5/mod.3.2. The results of the calculation and experiment are compared. The code properly simulate the analyzed transient.(author)

  20. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonella L., E-mail: lombardicosta@gmail.co [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Reis, Patricia Amelia L., E-mail: patricialire@yahoo.com.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Pereira, Claubia, E-mail: claubia@nuclear.ufmg.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Veloso, Maria Auxiliadora F., E-mail: dora@nuclear.ufmg.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN, Av. Antonio Carlos, 6627, Campus UFMG, Belo Horizonte (Brazil); Soares, Humberto V., E-mail: betovitor@ig.com.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2010-06-15

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  1. User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Forsmann, J. Hope [Idaho National Lab. (INL), Idaho Falls, ID (United States); Weaver, Walter L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.

  2. Development of LabVIEW web-based simulator for RELAP5

    International Nuclear Information System (INIS)

    This work presents the development of a LabVIEW web-based simulator using the output results of the best estimate nuclear system analysis code, RELAP5, for graphical user interfaces and web-casting. A numerical based model designed for natural circulation studies on the thermal hydraulic experimental facility called Natural Circulation Circuit, was developed with RELAP5 code. Specific output results from RELAP5 simulation are displayed in a user friendly graphical format. The temperatures are shown as a function of time in a XY graphic. Temperatures, levels and void fractions are displayed in color-coded scale which change in time on the graphical interface representing the circuit. An alarm is set for the case of onset boiling temperature occurrence at the heater outlet. This simulator allows an easy visual understanding of the thermal hydraulic circuit behavior. It can be shared, via Web, with researchers in any geographical location and, at the same time, it can be used in learning for distance educational purposes. In future work, this LabVIEW simulator will be coupled with RELAP5 code through dll's. Simultaneous graphical displaying and code calculations will be possible. Results are presented and discussed. (author)

  3. Animation model of Krsko nuclear power plant for RELAP5 calculations

    Energy Technology Data Exchange (ETDEWEB)

    Prosek, Andrej, E-mail: andrej.prosek@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Mavko, Borut [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Today most software applications, also in the nuclear field, come with a graphical user interface. The first graphical user interface for the RELAP5 thermal-hydraulic computer code was called the Nuclear Plant Analyzer (NPA). Later, Symbolic Nuclear Analysis Package (SNAP) was developed. The purpose of the present study was to develop SNAP animation model of Krsko nuclear power plant (NPP) for RELAP5 calculations with the aim to help analyze the results. In addition, the reference calculations for Krsko full scope simulator validation were performed with the latest RELAP5/MOD3.3 Patch 03 code and compared to previous RELAP5 versions to provide verified source data, needed to demonstrate animation model. In total six scenarios were analyzed: two scenarios of the small-break loss-of-coolant accident, two scenarios of the loss of main feedwater, a scenario of the anticipated transient without scram, and a scenario of the steam generator tube rupture. The use of SNAP for animation of Krsko nuclear power plant analyses showed several benefits, especially better understanding of the calculated physical phenomena and processes. It can be concluded that an animation tool was created, which enables to analyze very complex accident scenarios. The graphical surface helps keeping the overview and focusing on the main influences. Also, the use of such support tools to system codes may significantly contribute to better quality of safety analysis.

  4. Development of LabVIEW web-based simulator for RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Luiz A.; Torres, Walmir M.; Sabundjian, Gaiane; Andrade, Delvonei A.; Belchior Junior, Antonio; Umbehaun, Pedro E.; Conti, Thadeu N.; Mesquita, Roberto N. de; Masotti, Paulo H.F.; Angelo, Gabriel, E-mail: lamacedo@ipen.b, E-mail: wmtorres@ipen.b, E-mail: gdjian@ipen.b, E-mail: delvonei@ipen.b, E-mail: abelchior@ipen.b, E-mail: umbehaun@ipen.b, E-mail: tnconti@ipen.b, E-mail: rnavarro@ipen.b, E-mail: , E-mail: masotti@ipen.b, E-mail: gabriel.angelo@usp.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work presents the development of a LabVIEW web-based simulator using the output results of the best estimate nuclear system analysis code, RELAP5, for graphical user interfaces and web-casting. A numerical based model designed for natural circulation studies on the thermal hydraulic experimental facility called Natural Circulation Circuit, was developed with RELAP5 code. Specific output results from RELAP5 simulation are displayed in a user friendly graphical format. The temperatures are shown as a function of time in a XY graphic. Temperatures, levels and void fractions are displayed in color-coded scale which change in time on the graphical interface representing the circuit. An alarm is set for the case of onset boiling temperature occurrence at the heater outlet. This simulator allows an easy visual understanding of the thermal hydraulic circuit behavior. It can be shared, via Web, with researchers in any geographical location and, at the same time, it can be used in learning for distance educational purposes. In future work, this LabVIEW simulator will be coupled with RELAP5 code through dll's. Simultaneous graphical displaying and code calculations will be possible. Results are presented and discussed. (author)

  5. Recent Hydrodynamics Improvements to the RELAP5-3D Code

    International Nuclear Information System (INIS)

    The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model

  6. RELAP5 two-phase fluid model and numerical scheme for economic LWR system simulation

    International Nuclear Information System (INIS)

    The RELAP5 two-phase fluid model and the associated numerical scheme are summarized. The experience accrued in development of a fast running light water reactor system transient analysis code is reviewed and example of the code application are given

  7. Recent Hydrodynamics Improvements to the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Richard A. Riemke; Cliff B. Davis; Richard.R. Schultz

    2009-07-01

    The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model.

  8. Assessment of RELAP5/MOD3 with condensation experiment for pure steam condensation in a vercal tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Jae; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The film condensation models in RELAP5/MOD3.1 and RELAP5/MOD3.2 are assessed with the data of experiment performed in the scaled down condensation experimental facility with a single vertical tube of inner diameter of 46 mm in the range of pressure 0.1 {approx} 7.5 MPa for the PSCS(Passive Secondary Condenser System). Both MOD3.1 and MOD3.2 don`t shows any reliable predictions of the experimental data. The RELAP5/MOD3.1 overpredicts the heat transfer coefficients of experiment, whereas the RELAP5/MOD3.2 underpredicts those data. It is recommended that the film condensation model in RELAP5/MOD3.2 should be modified to have a larger heat transfer coefficient than those of the present model to give the reliable predictions. 7 refs., 6 figs., 1 tab. (Author)

  9. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  10. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

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

    International Nuclear Information System (INIS)

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

  12. Plant application uncertainty evaluation of LBLOCA analysis using RELAP5/MOD3/KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Chung, Bub Dong; Hwang, Tae Suk; Lee, Guy Hyung; Chang, Byung Hoon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the revised ECCS rule has been developed and this report describes the application of new REM to large break LOCA. A computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/ MOD3.1 was used as the best estimated code for the analysis and Kori unit 3 and 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by using Monte-Carlo sampler for each response surface. This study shows that plant application uncertainty can be quantified and demonstrates the applicability of the new realistic evaluation methodology. (Author) 29 refs., 40 figs., 8 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

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

  14. Heat Transfer Boundary Conditions in the RELAP5-3D Code

    International Nuclear Information System (INIS)

    The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones

  15. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Putney, J.M.; Preece, R.J. [National Power, Leatherhead (GB). Technology and Environment Centre

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  16. Heat Transfer Boundary Conditions in the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Richard A. Riemke; Cliff B. Davis; Richard R. Schultz

    2008-05-01

    The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.

  17. Analysis of postulated loss of coolant accidents on Brazilian Multipurpose Reactor using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Humberto Vitor; Costa, Antonella Lombardi; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Reis, Patricia Amelia de Lima, E-mail: hvs@cdtn.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: patricialire@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Aronne, Ivan Dionysio, E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-07-01

    The Brazilian Multipurpose Reactor (RMB) is currently being projected and several analyses are being carried out. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly with planar plates. RMB will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of RMB using the RELAP5 model and also three cases of loss of coolant accident (LOCA), in the reactor and service polls cooling system (RSPCS) inlet and two cases in the primary coolant system (PCS), inlet and outlet. In both cases the coolant pool level decreased until 7 m, keeping the core covered by water, but in different times. Natural circulation mode was established in the reactor pool and consequently the decay heat was removed keeping the integrity of the fuel elements. Keywords: Research reactor, LOCA, RELAP5. (author)

  18. Evaluation of validity of the RELAP5/MOD3 flow regime map for horizontal tubes

    International Nuclear Information System (INIS)

    RELAP5/MOD3 code was developed for western type power water reactors with vertical steam generators. Thus, this code should be validated also for VVER design with horizontal steam generators. The validation work, which has been started in Lappeenranta University of Technology (LUT), has already shown some weaknesses of the code. For example the flow inside a steam generator horizontal tube in some accident cases is not correctly modelled by the code. It may be the result of erroneous prediction of the flow regime. The aim of the study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal tubes. (18 refs.)

  19. Krsko核电厂RELAP5仿真分析

    Institute of Scientific and Technical Information of China (English)

    Andrej Prsek; Borut Mavko; 于明锐(译)

    2011-01-01

    现在大多数应用软件都采用图形用户界面,首先把这项技术应用于RELAP5计算机模拟中的是核电站分析软件(NPA),后来开发出了图像核子分析软件(SNAP)。本研究的目的是开发电站SNAP动画模型,用RELAP5进行模拟计算并分析结果。此外,采用最新的RELAP5/MOD3.3 Patch03进行全范围模拟机验证的参考计算,和以前RELAP5版本相比提供了源数据的验证,但新版本需要建立动画模型。本研究分析了下面几种事故工况:小破口失水事故;主给水丧失事故;未能紧急停堆的预计瞬变;蒸汽发生器传热管破裂事故。使用SNAP分析系统动画模型能更好地了解计算过程和物理现象。由此可知,创建一个成功的模型可以帮助分析非常复杂的事故情况,而且有助于分析主要影响因素和保持大局观。此外,利用这些工具对提高系统代码安全分析的质量也有很大帮助。

  20. RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

    OpenAIRE

    Borut Mavko; Andrej Prošek

    2010-01-01

    To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA). The current PSA standard recommends the use of best-estimate codes. The purpose of the study was to estimate the operator action success criteria time windows in scenarios in which the human actions are supplement to safety systems actuations, needed for updated human reliability analysis (HRA). For calculations the RELAP5/MOD3.3 best ...

  1. Recent Heat Transfer Improvements to the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard A; Davis, Cliff B; Oh, Chang

    2007-05-01

    The heat transfer section of the RELAP5-3D computer program has been recently improved. The improvements are as follows: (1) the general cladding rupture model was modified (more than one heat structure segment connected to the hydrodynamic volume and heat structure geometry’s internal gap pressure), (2) the cladding rupture model was modified for reflood, and (3) the heat transfer minor edits/plots were extended to include radiation/enclosure heat flux and generation (internal heat source).

  2. Assessment of the code RELAP5/MOD2 against loss of feedwater without scram

    International Nuclear Information System (INIS)

    The integral effect test L9-3 (loss of feedwater without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transients in pressurized water reactors. The code proved suitable for analysis of this kind of transients. Some conclusions of relevance to simulation of anticipated transients without scram scenarios with forced circulation could be drawn. (orig.)

  3. Assessment of the code RELAP5/MOD2 against loss of feedwater without scram

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-02-01

    The integral effect test L9-3 (loss of feedwater without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transients in pressurized water reactors. The code proved suitable for analysis of this kind of transients. Some conclusions of relevance to simulation of anticipated transients without scram scenarios with forced circulation could be drawn. (orig.).

  4. Assessment of RELAP5/CANDU+ code for regulatory auditing analysis of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Kim, Hho Jung; Yang, Chae Yong

    2001-12-15

    The objectives of this study are to undertake the verification and validation of RELAP5/CANDU+ code, which is developed in this project, by simulating the B8711 test of RD-14 facility, and to examine the properties of this code by doing the sensitivity analysis for experimental prediction modes about thermal-hydraulics phenomena in CANDU reactor systems added to this code. The B8711 test was an experiment of a 45% ROH break for simulating large LOCA. Also, in this study, the methods for making input cards related to CANDU options are described, so that some users can use the RELAP5/CANDU+ code with easy. RELAP/CANDU+ code can choose the options of Henry-Fauske mode, Ransom-Trapp model, and Moody model for prediction of the critical mass flow. It is examined that Henry-Fauske model and Ransom-Trapp model are considered properly, but Moody model is still required to be improved. Heat transfer correlations available in RELAP5/CANDU+ code for CANDU-type reactors are a horizontal stratified model, a fuel heat-up model and D2O/H2O CHF correlations, and these models take an important role to improve the predictability of the experimental procedures. It is concluded that RELAP5/CANDU+ code is useful for the auditing of the accident analysis of CANDU reactors, and the results of the sensitivity analysis for thermal-hydraulic models examined in this study are valuable for the actual auditing of real CANDU-type power plants.

  5. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, N.; Khedr, A. [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt); D' Auria, F.D. [Pisa Univ. (Italy). Facolta di Ingegneria

    2015-12-15

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

  6. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    International Nuclear Information System (INIS)

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

  7. Triangular coupling of SCDAP/Lower Head/RELAP5 using parallel virtual machine

    International Nuclear Information System (INIS)

    The present integrated SCDAP/RELAP5 MOD3.2 computer code is used for the simulation of reactor coolant system thermal-hydraulic (T/H) response and core damage progression. Due to the size of the code (∼150,000 lines), maintenance and upgrades present a significant burden. For the user, the code structure and data communication is difficult to understand, thus making it difficult to track down problems that arise during severe accident analyses. Adding to this difficulty is the fact that typical simulations can take anywhere from a few days to several weeks to complete. Therefore, in an effort to reduce the maintenance burden and enhance code performance, a change in the code structure was required. The U.S. Nuclear Regulatory Commission (USNRC) and the Swiss Federal Nuclear Safety Inspectorate (HSK), with technical assistance from SCIENTECH, Inc. and the Idaho Engineering and Environmental Laboratory (INEEL) initiated a concept of a triangular coupling of sub-modules within the SCDAP/RELAP5 MOD3.2 code: SCDAP, Lower Head (LOWHD), and RELAP5. The goal was to cleanly separate the SCDAP, LOWHD, and RELAP5 modules from the integrated SCDAP/RELAP5 code and then link these code modules using the Parallel Virtual Machine (PVM) software package. This triangular coupling using PVM is now successfully completed, and will facilitate the coupling of SCDAP/LOWHD with the new generation of thermal hydraulic codes, such as TRAC-M. With regard to parallel performance, this work is only a first step towards obtaining significant performance improvements. Further efforts will be required to fully realize these parallel performance gains. These are the further goals the USNRC and HSK propose to follow in the near future. The coupled code was verified by comparing results with those of the original merged code (i.e. the code as it existed prior to separating the three codes). First, three relatively simple developmental assessment cases were run, including a bundle boil

  8. An analysis of MB-2 100% steam line break test T-2013 using RELAP5/MOD2

    International Nuclear Information System (INIS)

    This report presents RELAP5/MOD2 calculations of the 100% steam line break test T-2013 performed on the Westinghouse Model Boiler-2 facility (MB-2). The input deck uses a noding structure typical of what would be used for an integral rig or full plant study using the RELAP5/MOD2 code. Sensitivity calculations were performed for the break junction discharge coefficient and the separator drain line loss coefficient. (author)

  9. Pre-test of the KYLIN-II thermal-hydraulics mixed circulation LBE loop using RELAP5

    International Nuclear Information System (INIS)

    To investigate the behavior of lead bismuth eutectic (LBE) as coolant in China LEAd-based Research Reactor, Institute of Nuclear Energy Safety Institute (INEST), Chinese Academy of Sciences has built a multi-functional LBE experiment facility KYLIN-II. Mixed circulation loop, which is one of the KYLIN-II thermal-hydraulics loops, has the capability to drive the flowing LBE in different ways such as pump, gas lift and temperature difference (natural circulation). In this contribution, preliminary numerical simulations in support of the operation and experiment of KYLIN-II thermal-hydraulics mixed circulation LBE loop have been carried out and the obtained results have been studied. The RELAP5 Mod4.0 with LBE model has been utilized. Pre-test analysis showed the LBE circulation capability can reach the object under several driven patterns. The maximum velocity in fuel pin bundles can be larger than 0.15 m/s for natural circulation, 0.5 m/s for gas enhanced circulation, and 2 m/s for pump driven circulation. (author)

  10. Study of the Relap5/mod3.2 wall heat flux partitioning model

    Energy Technology Data Exchange (ETDEWEB)

    Hari, S.; Hassan, Y.A. [Texas A and M University, Dept. of Nuclear Engineering, College Station, TX (United States)

    2001-07-01

    The performance of the subcooled boiling model adapted in RELAP5/MOD3.2 computer code has been assessed in detail for low-pressure conditions and it has been found that the void fraction profile is under-predicted. In general, any subcooled boiling model is composed of individual sub-models that account for the different physical mechanism that govern the overall process, as the wall vapor generation, interfacial shear and condensation etc. The wall heat flux partitioning model is one of the important sub-models that is a constituent of any subcooled boiling model. The function of this model is to apportion the wall heat flux to the different components (as the single/two phase fluid or bubble), as the case may be, in a two-phase flow-boiling scenario adjacent to a heated wall. The ''pumping factor'' approach is generally followed by most of the wall heat flux partitioning models, for partitioning the wall heat flux. In this work, the wall heat flux partitioning model of RELAP5/MOD3.2 computer code is studied; in particular, the ''pumping factor'' formulation in the present code version is assessed for its performance under low-pressure conditions. In addition, three different ''pumping factor'' formulations available in the literature have been introduced into the RELAP5/MOD3.2 code. Simulations of two low-pressure subcooled flow boiling experiments were performed with the refined code versions to determine the appropriate pumping factor to be used under these conditions. (author)

  11. An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, B.D.; Parlatan, Y.; Slovik, G.C. [and others

    1995-09-01

    RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {plus_minus}5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes.

  12. Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

  13. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

    International Nuclear Information System (INIS)

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  14. Methodology, status, and plans for development and assessment of the RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, G.W.; Riemke, R.A. [Idaho National Engineering Laboratory, Idaho Falls, ID (United States)

    1997-07-01

    RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

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

    International Nuclear Information System (INIS)

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

  16. RELAP5/MOD3 code manual: User's guide and input requirements. Volume 2

    International Nuclear Information System (INIS)

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume II contains detailed instructions for code application and input data preparation

  17. Steady state and transient analyses of MNSR reactor using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, E.; Khorsandi, Jamshid [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research School; Tashakor, S. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research School; Islamic Azad Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-03-15

    Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.

  18. Developmental assessment of the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    International Nuclear Information System (INIS)

    This report documents ten developmental assessment problems which were used to test the multidimensional component in RELAP5/MOD2.5, Version 3w. The problems chosen were a rigid body rotation problem, a pure radial symmetric flow problem, an r-θ symmetric flow problem, a fall problem, a rest problem, a basic one-dimensional flow test problem, a gravity wave problem, a tank draining problem, a flow through the center problem, and coverage analysis using PIXIE. The multidimensional code calculations are compared to analytical solutions and one-dimensional code calculations. The discussion section of each problem contains information relative to the code's ability to simulate these problems

  19. Improvements to the RELAP5-3D Nearly-Implicit Numerical Scheme

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved with regard to its nearly-implicit numerical scheme for two phase flow and single-phase flow. Changes were made to the nearly-implicit numerical scheme finite difference momentum equations as follows: (1) added the velocity flip-flop mass/energy error mitigation logic, (2) added the modified Henry-Fauske choking model, (3) used the new time void fraction in the horizontal stratification force terms and gravity head, and (4) used an implicit form of the artificial viscosity. The code modifications allow the nearly-implicit numerical scheme to be more implicit and lead to enhanced numerical stability

  20. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point

  1. RELAP5-3D code for supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)

  2. Improvements to the RELAP5-3D Nearly-Implicit Numerical Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Richard A. Riemke; Walter L. Weaver; RIchard R. Schultz

    2005-05-01

    The RELAP5-3D computer program has been improved with regard to its nearly-implicit numerical scheme for twophase flow and single-phase flow. Changes were made to the nearly-implicit numerical scheme finite difference momentum equations as follows: (1) added the velocity flip-flop mass/energy error mitigation logic, (2) added the modified Henry-Fauske choking model, (3) used the new time void fraction in the horizontal stratification force terms and gravity head, and (4) used an implicit form of the artificial viscosity. The code modifications allow the nearly-implicit numerical scheme to be more implicit and lead to enhanced numerical stability.

  3. Steady state and transient analyses of MNSR reactor using RELAP5 code

    International Nuclear Information System (INIS)

    Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.

  4. Simulation of the LOFT L9-4 experiment with the code RELAP5/MOD2

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-02-01

    The integral effect test L9-4 (loss of off-site power without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transient in pressurized water reactors. The code was qualified for analysis of the thermal-hydraulics and kinetics associated to this kind of sequences. Some conclusions concerning simulation of anticipated transients without scram scenarios under natural circulation and axial power profile redistribution in power reactors are derived. (orig.).

  5. Simulation of the LOFT L9-4 experiment with the code RELAP5/MOD2

    International Nuclear Information System (INIS)

    The integral effect test L9-4 (loss of off-site power without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transient in pressurized water reactors. The code was qualified for analysis of the thermal-hydraulics and kinetics associated to this kind of sequences. Some conclusions concerning simulation of anticipated transients without scram scenarios under natural circulation and axial power profile redistribution in power reactors are derived. (orig.)

  6. RELAP5/MOD3 code manual: User`s guide and input requirements. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume II contains detailed instructions for code application and input data preparation.

  7. Implementation of a new bubbly-slug interphase drag model in RELAP5/MOD2

    International Nuclear Information System (INIS)

    The implementation of a new bubbly-slug interphase drag model in the RELAP5/MOD2 code is described. The model is based on the determination of an effective interphase drag coefficient from a set of best-estimate void fraction correlations covering the full range of geometries and flow conditions encountered in PWR safety analysis. Calculations are reported which show that the new model leads to a much better prediction of void fraction profile for low flows in rod bundles than the standard model. Further work is necessary to derive a model formulation which can be guaranteed to produce physical drag coefficients in all flow situations. (author)

  8. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    OpenAIRE

    2015-01-01

    Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed tha...

  9. 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.

  10. SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported

  11. RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2010-01-01

    Full Text Available To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA. The current PSA standard recommends the use of best-estimate codes. The purpose of the study was to estimate the operator action success criteria time windows in scenarios in which the human actions are supplement to safety systems actuations, needed for updated human reliability analysis (HRA. For calculations the RELAP5/MOD3.3 best estimate thermal-hydraulic computer code and the qualified RELAP5 input model representing a two-loop pressurized water reactor, Westinghouse type, were used. The results of deterministic safety analysis were examined what is the latest time to perform the operator action and still satisfy the safety criteria. The results showed that uncertainty analysis of realistic calculation in general is not needed for human reliability analysis when additional time is available and/or the event is not significant contributor to the risk.

  12. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

  13. Proposals for improving interphase drag modelling for the bubbly and slug regimes in RELAP5

    International Nuclear Information System (INIS)

    The proposal is put forward that the effective interphase drag coefficient for the bubbly and slug regimes in RELAP5 should be calculated using best-estimate void fraction correlations. It is argued that this will lead to improvements in the code's modelling of interphase drag and evidence is given to corroborate this. The need for such improvements has been prompted by the poor performance of the current models in simulating rod bundle experiments. There is also concern that the models do not account for profile slip effects, which could be important in a variety of geometries, and that the slug flow equations may not be appropriate for large diameter vertical pipes. To support the proposal, a set of void fraction correlations is identified which is believed to cover the full range of geometries and flow conditions encountered in PWR safety analysis including the analysis of small-scale experimental facilities. This set is selected from a detailed appraisal of the most appropriate correlations found in the literature which takes account of comparisons with experimental data and physical considerations. This Report forms part of the UK's commitment to the ICAP Code Improvement Plan. The recommendations will now be implemented in a development version of RELAP5/MOD3 and a preliminary assessment made. The interphase drag models used in the annular-mist regime will also be examined and, if necessary, appropriate improvements will be proposed. (author)

  14. SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K.; Johnsen, E.C. [eds.; Allison, C.M. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported.

  15. Analysis of the OECD-LOFT International Standard Problem 31 using SCDAP/RELAP5/MOD3

    International Nuclear Information System (INIS)

    The CORA-13 bundle heating and melting experiment performed at the Kernforechungszentrum, Karlaruhe, (KfK) was analyzed at the Idaho National Engineering Laboratory (INEL) using SCDAP/RELAP5/MOD3. This analysis was part of a systematic assessment of SCDAP/RELAP5/MOD3 for the US Nuclear Regulatory Commission to (a) evaluate the variances between calculated and observed behavior, (b) identify outstanding modeling deficiencies, and (c) to evaluate the impact of ongoing modeling improvements. A brief discussion of the CORA-13 experiment including a description of the facility, important test conditions, and comparisons with other CORA experimental conditions and results is provided in this report. This report describes the results of the SCDAP/RELAPS/MOD3 analysis including a description of the SCDAP/RELAPS model of the facility, base case results, sensitivity results, and a comparison with other SCDAP/RELAP5/MOD3 code-to-data comparisons

  16. Assessment of RELAP5/MOD3.2 with condensation experiment in the presence of noncondensables in a vertical tube

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, and simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube. 10 refs., 6 figs. (Author)

  17. RELAP5-3D version 4.0.3: installation and tests for applications to space reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, Paulo D.C.; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine N.F., E-mail: plobo.a@uol.com.br, E-mail: fbraz@ieav.cta.br, E-mail: eduardo@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    To attend the TERRA project (Tecnologia de Reatores Rapidos Avancados), currently conducted by the Nuclear Energy Division (ENU) of the IEAv, this work presents the RELAP5-3D, Version 4.0.3, prepared in July 12, 2012, also known as r3d403is, received recently by the IEAv from the Idaho National Laboratory (INL). This version of RELAP5-3D is configured for the International User Group source Code Group and is developed and maintained at the INL for the US Department of Energy. RELAP5-3D, the latest in the series of RELAP5 codes, is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. Enhancements include all features and models previously available in the ATHENA configuration version of the code which are as follows: addition of new work fluids and a magneto-hydrodynamic mode. Following the instructions from the README file, the RELAP5-3D, version 4.0.3 was installed creating the necessaries subdirectories, by using the LINUX platform and applying both Intel Fortran 95 and C-language compilers. Many input examples were executed and the same results were observed as compared to the received documentation. A sample of the Edwards-O'Brien test was evaluated to verify if the code could simulate a LOCA type accident properly. The test executed by the RELAP5-3D demonstrated good agreement with test data including a new output involving the mass flow during the test. (author)

  18. RELAP5-3D Developmental Assessment: Comparison of Version 4.2.1i on Linux and Windows

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    2014-06-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.2i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

  19. RELAP5-3D Developmental Assessment. Comparison of Version 4.3.4i on Linux and Windows

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.3i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

  20. PUMA-PCCS separate effect tests and RELAP5 code evaluation in PUMA

    Science.gov (United States)

    Choi, Sung Won

    One of the key areas in the design of advanced nuclear reactors is to develop a reliable Passive Containment Cooling System (PCCS). The purpose of the current work is to better understand the condensation phenomena in PCCS for the downward co-current flow of a steam/air mixture through condenser tube bundles during the three PCCS operational modes, namely the bypass mode, the cyclic venting mode and the long-term cooling mode. A series of unique separate-effect PCCS test data were obtained for condensation heat transfer in the PCCS heat exchangers of the PUMA (Purdue University Multidimensional Integral Test Assembly) facility under a task sponsored by the U.S. Nuclear Regulatory Commission. Test conditions includes bypass mode, cyclic venting mode and long term mode, covering a wide range of Loss of Coolant Accident(LOCA) conditions with a parameters of pressure, mass flow rate, noncondensable(NC) gases, and PCCS pool water level. The parametric effect studies and a further validation of the PUMA-PCCS separate effect test data were performed. The evaluation of a best estimate system code (RELAP5/MOD3.3) was performed by using unique PUMA-PCCS separate effects data and PUMA-Main Steam Line Break (MSLB) integral test (1998). Through a sensitivity studies of nodalization method and physical models on the MSLB test simulations, deficiencies in RELAP5/MOD3.3 code were found as follows: (1) over prediction of heat removal rate by condensation models, (2) overestimation of SP heat transfer through the horizontal venting line and thermal stratification distortion, (3) underestimation of NC gas effects in PCCS by the distortion of cyclic venting phenomena and (4) overestimation of the DW and SP wall condensation. The improvement for the code calculation predictions could be obtained by removing the RELAP5/MOD3.3 code deficient factors in the PUMA MSLB integral test simulation. The unique PCCS NC gas venting visualizations were obtained according to various PCCS inlet NC

  1. Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Cliff B. Davis

    2010-09-01

    RELAP5-3D interpolates to obtain thermodynamic properties for use in its internal calculations. The accuracy of the interpolation was determined for the original steam tables currently used by the code. This accuracy evaluation showed that the original steam tables are generally detailed enough to allow reasonably accurate interpolations in most areas needed for typical analyses of nuclear reactors cooled by light water. However, there were some regions in which the original steam tables were judged to not provide acceptable accurate results. Revised steam tables were created that used a finer thermodynamic mesh between 4 and 21 MPa and 530 and 640 K. The revised steam tables solved most of the problems observed with the original steam tables. The accuracies of the original and revised steam tables were compared throughout the thermodynamic grid.

  2. TMI-2 analysis using SCDAP/RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K.; Polkinghorne, S.T.; Siefken, L.J.; Allison, C.M.; Dobbe, C.A.

    1994-11-01

    SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial conditions, boundary conditions, and the results from the best-estimate simulation of Phases 1 to 4 of the TMI-2 accident as well as the results from several sensitivity calculations.

  3. Vectrorization of the LWR transient analysis code RELAP5/MOD2/CYCLE36

    International Nuclear Information System (INIS)

    The LWR transient analysis code RELAP5/MOD2/CYCLE36 has been vectorized. The performance of the vectorized code in vector mode of the FACOM VP-100 was 3.5 times higher than that of original one in scalar mode for a real scale transient calculation. Subroutines for heat conduction were vectorized in terms of heat structures and heat meshes, while those for hydrodynamics were vectorized in terms of volumes and junctions. In this report, the vectorization method used for each of subroutines and its vectorization effect are described. A reduction of the overhead of bit operation functions which is introduced when the code was converted from the CDC to the FACOM is also described. (author)

  4. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    International Nuclear Information System (INIS)

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  5. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    Directory of Open Access Journals (Sweden)

    Patrícia A. L. Reis

    2015-01-01

    Full Text Available Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

  6. Interpretation of TRIGA reactivity transients with RELAP5/PARCS coupled-code

    International Nuclear Information System (INIS)

    In the frame of future experiments to carried out upon TRIGA reactors, which aim to verify the real feasibility of the ADS (Accelerator Driven System) concept, it is essential to build a numerical tool able to simulate the dynamic behaviour of the reactor in subcritical configuration. This model developed to support the design of subcritical experiments and the safety analysis of the reactor, as a first step has to be assessed against the experimental data available for the critical reactor. To this purpose the thermal-hydraulic/ neutronic numerical model based on the RELAP5/PARCS coupled-code is been tested against the experimental reactivity transients conducted on the RC1-TRIGA reactor at the ENEA Casaccia Research Center in forecast of the TRADE (TRIGA Accelerator Driven Experiment) subcritical experience. The results of the calculations already performed show a qualitative good agreement with the experimental data and allow to address the future developments and improvements of the numerical model. (authors)

  7. Evaluation and assessment of reflooding models in RELAP5/Mod2.5 and RELAP5/Mod3 codes using Lehigh University and PSI-Neptun bundle experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Sencar, M.; Aksan, N. [Paul Scherrer Institute, Villigen (Switzerland)

    1995-09-01

    An extensive analysis and assessment work on reflooding models of RELAP5/Mod2.5 and, RELAP5/Mod3/v5m5 and RELAP/Mod3/v7j have been performed. Experimental data from LehighUniversityv. and PSI-NEPTUN bundle reflooding experiments have been used for the assessment, since both of these tests cover a broad range of initial conditions. Within the range of these initial conditions, it was tried to identify their separate impacts on the calculated results. A total of six Lehigh University reflooding bundle tests and two PSI-NEPTUN tests with bounding initial conditions are selected for the analysis. Detailed nodalisation studies both for hydraulic and conduction heat transfer were done. On the basis of the results obtained from these cases, a base nodalisation scheme was established. All the other analysis work was performed by using this base nodalisation. RELAP5/Mod2.5 results do not change with renodalisation but RELAP5/Mod3 results are more sensitive to renodalisation. The results of RELAP5/Mod2.5 versions show very large deviations from the used experimental data. These results indicate that some of the phenomenology of the events occurring during the reflooding could not be identified. In the paper, detailed discussions on the main reasons of the deviations from the experimental data will be presented. Since, the results and findings of this study are meant to be a developmental aid, some recommendations have been drawn and some of these have already been implemented at PSI with promising results.

  8. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  9. Conversion tool for the LWR transient analysis code RELAP5 from the CDC version to the FACOM version

    International Nuclear Information System (INIS)

    The LWR transient analysis code RELAP5 has been developed on the CDC-CYBER 176 at Idaho National Engineering Laboratory (INEL), the RELAP5 code has been often updated in order to extend the analyzing model and correct the errors. At Japan Atomic Energy Research Institute the code has been converted from the CDC version to the FACOM version and the converted code has been used. The conversion is the task which consumes a lot of time, because the code is large and there is the difference between CDC's machines and FACOM's ones. In order to convert the RELAP5 code automatically, the software tool has been developed. By using this tools the efficiency for converting the RELAP5 code has been improved. Productivity of the conversion is increased about 2.0 to 2.6 times by the tools in comparison with in manual. The procedure of conversion by using the tools and the option parameters of each tool are described. (author)

  10. Theory and input requirements for the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    International Nuclear Information System (INIS)

    This report documents the theory and input requirements for the multidimensional component in RELAP5/MOD2.5, Version 3w. The equations in Cartesian and cylindrical coordinates are presented as well as the shallow water terms. The implementation of these equations is then discussed. Finally, the constitutive models and input requirements are then described

  11. Modeling a Printed Circuit Heat Exchanger with RELAP5-3D for the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-01

    The main purpose of this report is to design a printed circuit heat exchanger (PCHE) for the Next Generation Nuclear Plant and carry out Loss of Coolant Accident (LOCA) simulation using RELAP5-3D. Helium was chosen as the coolant in the primary and secondary sides of the heat exchanger. The design of PCHE is critical for the LOCA simulations. For purposes of simplicity, a straight channel configuration was assumed. A parallel intermediate heat exchanger configuration was assumed for the RELAP5 model design. The RELAP5 modeling also required the semicircular channels in the heat exchanger to be mapped to rectangular channels. The initial RELAP5 run outputs steady state conditions which were then compared to the heat exchanger performance theory to ensure accurate design is being simulated. An exponential loss of pressure transient was simulated. This LOCA describes a loss of coolant pressure in the primary side over a 20 second time period. The results for the simulation indicate that heat is initially transferred from the primary loop to the secondary loop, but after the loss of pressure occurs, heat transfers from the secondary loop to the primary loop.

  12. Modification of blowdown heat transfer models for RELAP5-3D in accordance with appendix K of 10CFR50

    Energy Technology Data Exchange (ETDEWEB)

    Chin-Jang, Chang; Liang, T.K.S. [Nuclear Engineering Div. Institute of Nuclear Energy Research, Lung-Tan, Taiwan (China); Huan-Jen, Hung; Wang, L.C. [Power Research Institute, Taiwan Power Company (China)

    2001-07-01

    The objective of this paper is to implement the blowdown heat transfer models accepted by Appendix K of 10CFR50 into RELAP5-3D and to rename it as RELAP5-3D/K. Modifications of critical heat flux (CHF) model, post-CHF model, and the heat transfer logic for nucleate and transition boiling lockout are included. Also the assessments against separate-effect experiments were evaluated for RELAP 5-3D/K. From calculation results, the conservative predictions of surface peak temperatures using RELAP5-3D/K are obtained. It demonstrated that the blowdown heat transfer models were successfully modified and implemented into RELAP5-3D in accordance with Appendix K of 10CFR50. (authors)

  13. Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Harrell, J.R. [ENERCON Services, Inc., Atlanta, GA (United States); Fuller, R.W. [Entergy Operations, Inc., Port Gibson, MS (United States)

    1996-07-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. The original design and testing requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  14. Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

    Energy Technology Data Exchange (ETDEWEB)

    Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane, E-mail: delvonei@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

  15. Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

    International Nuclear Information System (INIS)

    Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

  16. Uncertainty and sensitivity analyses of the Kozloduy pump trip test by coupled RELAP5/PARCS code

    Energy Technology Data Exchange (ETDEWEB)

    Bousbia Salah, A. [Pisa Univ., Facolta di Ingegneria, DIMNP (Italy); Kliem, S.; Rohde, U. [Forschungszentrum Rossendorf (FZR) (Germany)

    2005-07-01

    The modeling of complex transients in Nuclear Power Plants (NPP) remains a challenging topic for Best Estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulating and more precise consideration of multidimensional effects under complex transients in NPPs. In the current paper a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data and the comparison study shows good agreements between the calculations and the global kinetic and thermal-hydraulic aspects observed experimentally. It appears that at steady state level, the simulation errors are mainly due to the absence of ADF correction and to the model used for evaluating the Doppler feedback effect. During the transient, the discrepancies are mainly due to the combined effect of uncertain parameters related to the measurement of control rod course, and the estimation of the Doppler effect.

  17. Total loss of CNA1 steam generators feed water simulated with RELAP5/MOD3

    International Nuclear Information System (INIS)

    The results of the calculations are presented carried out by utilizing the code RELAP5/MOD3, upon the basis of the postulated initial event of total loss of feed water to the two steam generators in the nuclear power plant Atucha 1, CNA1. The evolution of the installation systems during the transient was analyzed in different conditions of availability: condenser, relief valve and safety valves in the secondary system, safety valves in the primary system and system of long-term subsequent cooling. Located in the primary and secondary systems of the installation they turn out to be prominent in this event. Upon this basis the sequences of possible evolution were calculated and those that would conduct the system toward the setting called 'damage to the core' were determined. Also those in which would arrive to a state of 'safe shutdown' were determined. These results were utilized in the verification of the tree of events utilized in the Final Report of the Probabilistic Safety Analysis for the sequence of event T9, made from calculations carried out with the code DINETZ. From this compare some differences were determined and are presented in the modified version of tree of events. (author)

  18. Reformulation RELAP5-3D in FORTRAN 95 and Results

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2010-08-01

    RELAP5-3D is a nuclear power plant code used worldwide for safety analysis, design, and operator training. In keeping with ongoing developments in the computing industry, we have re-architected the code in the FORTRAN 95 language, the current, fully-available, FORTRAN language. These changes include a complete reworking of the database and conversion of the source code to take advantage of new constructs. The improvements and impacts to the code are manifold. It is a completely machine-independent code that produces machine independent fluid property and plot files and expands to the exact size needed to accommodate the user’s input. Runtime is generally better for larger input models. Other impacts of code conversion are improved code readability, reduced maintenance and development time, increased adaptability to new computing platforms, and increased code longevity. The conversion methodology, code improvements and testing upgrades are presented in a manner that will be useful to future conversion projects for other such large codes. Comparison between the pre- and post-conversion code are made on the basis of code metrics and code performance.

  19. RELAP5/MOD2 shutdown thermal-hydraulic analysis for NPP Krsko

    International Nuclear Information System (INIS)

    The purpose of the previous simplified and very conservative shutdown thermalhydraulic analysis [2] was to determine the ''time to boiling'' and ''time to core uncover'' in the event of a loss of Decay Heat Removal (DHR). The various Reactor Coolant System (RCS) levels and pressure boundaries to predict necessary containment closure time (CCT) were assumed. RCS heatup rates, the required makeup rates (required flow into RCS to prevent boiling) and required RCS mass flows to maintain subcooling were derived. Main purpose of that analysis was to provide input to overall analysis of plant safety during shutdown. The previous analysis [2] was performed with the most conservative assumptions like thermal isolation of reactor coolant loops from core, etc. The loss of water through pressurizer during a loss of DHR on midle loop operation was not considered. This paper presents re-analysis of chosen case (Pressurizer Manway Open - PMO) by a simplified analysis with best-estimate RELAP5/MOD3.2.2 gamma calculation taking into account reference initial and boundary conditions and modeled geometrical correspondence between plant zones and control volumes. There is also investigation of possible discharging flow via PMO is considered. (author)

  20. SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made

  1. Preliminary use of RELAP5 Code with Internal Assessment of Uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F.; Giannotti, W. [Dipartimento di Costruzioni Meccaniche e Nucleari, University of Pisa, Pisa (Italy)

    1999-07-01

    The present work deals with the preliminary use of the RELAP5 code with the internal assessment of the uncertainty. The activity has been started as a follow up of the research leading to the proposal and applicability of UMAE (Uncertainty Methodology base on Accuracy Extrapolation) uncertainty methodology. The concept of hypercube to characterize the status of a LWR plant during any transient and the assumption of uncertainty connected with the hypercubes are adopted: the CIAU (Code with Internal Assessment of Uncertainty) method has been set up. In the frame of the activity of CIAU development, matrix and a vector for uncertainty have been obtained; three type of matrixes and vectors are utilized: 1. matrix and vector deriving from tests results qualified by UMAE conditions; 2. matrix and vector deriving from tests results obtained from several sources (UMAE et not UMAE qualified) 3. matrix and vector used only for testing purposes (not derived from a physical process) are derived to evaluate the uncertainty connected with a generic time trend calculated by the code. The present paper describes an application of matrix and vector at item 3. (author)

  2. Analysis of void fraction in single channel using TRACE, MARS-KS, and RELAP5

    International Nuclear Information System (INIS)

    The accurate prediction of the void fraction is one of the most important factors in subchannel analyses. In general, the subchannel analysis has been conducted by means of dedicated subchannel analysis codes. However, the recent development of system codes with advanced two-phase flow model and three-dimensional components has helped more accurate and precise prediction of multiphase phenomena in nuclear reactor systems. This paper aims at evaluating the applicability of three different system codes to the prediction of the void fraction based on NUPEC experiment results employed for OECD/NRC PSBT benchmark. As a first step to the full scope analysis, analysis results for single channel experiments by using TRACE 5.0 Patch 3, MARS-KS 1.3, and RELAP5/MOD3.3 Patch 4 are presented in this paper. The result indicates that all codes slightly over-predict the void fraction compared to the experimental results in general and no significant discrepancies between the codes are observed. (author)

  3. RELAP5 analysis of integral test of the loss-of-RHR event during the mid-loop operation

    International Nuclear Information System (INIS)

    We are developing a statistical safety evaluation method using the RELAP5/MOD3.2 code for the loss-of-RHR (Residual Heat Removal) event during the mid-loop operation. To confirm the code prediction performance for gravity injection which is a one of the mitigation measures for this event, the Bethsy 6.9a test was analyzed using RELAP5/MOD3.2. In the analytical results, water mass flow rate into the pressurizer in the early period of the transient event was overpredicted. But water mass flow rate into the pressurizer was able to be decreased by artificially giving the circulation flow between core and the core bypass region. (author)

  4. Application of RELAP5/MOD3.1 code to the LOFT test L3-6

    Energy Technology Data Exchange (ETDEWEB)

    Pylev, S.S.; Roginskaja, V.L.

    1998-02-01

    A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs.

  5. Implementation of reactor safety analysis code RELAP5/MOD3 and its vectorization on supercomputer FACOM VP2600

    International Nuclear Information System (INIS)

    RELAP5/MOD3 is an advanced reactor safety analysis code developed at Idaho National Engineering Laboratory (INEL) under the sponsorship of USNRC. The code simulates thermohydraulic phenomena involved in loss of coolant accidents in pressurized water reactors. The code has been introduced into JAERI as a part of the technical exchange between the JAERI and USNRC under the ROSA-IV Program. First, the conversion to FACOM (= FUJITSU) M-780 version was carried out based on the IBM version extracted from the original INEL RELAP5/MOD3 source code. Next, the FACOM version has been vectorized for efficient use of new supercomputer FACOM VP2600 at JAERI. The computing speed of vectorized version is about three times faster than the scalar. The present vectorization ratio is 78%. In this report, both the implementation and vectorization methods on the FACOM computers are described. (author)

  6. Analysis of an extreme loss of coolant in the IPR-R1 TRIGA reactor using a RELAP5 model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia Amelia de Lima; Costa, Antonella Lombardi; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Soares, Humberto Vitor, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: hvs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Mesquita, Amir Zacharias, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-07-01

    The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPRR1 TRIGA have been validated for steady state and transient situations. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits. (author)

  7. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    International Nuclear Information System (INIS)

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components

  8. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

    1995-12-31

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

  9. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    Science.gov (United States)

    Sabundjian, Gaianê; Andrade, Delvonei A.; Belchior, Antonio, Jr.; da Silva Rocha, Marcelo; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; de Souza Lima, Ana Cecília

    2013-05-01

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm2, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  10. An assessment of the annular flow transition criteria and interphase friction models in RELAP5/MOD2

    International Nuclear Information System (INIS)

    An assessment of the annular flow transition criteria and interphase friction models for two-phase flow in tubes used in RELAP5/MOD2 code is described. The assessment examines the theoretical bases for the criteria and models and considers the results of comparisons with experimental data. Several deficiencies in the transition criteria are identified and appropriate improvements proposed. The interphase friction models are found to be adequate for PWR analyses. (author)

  11. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  12. SCDAP/RELAP5分析UO2-Zr板型元件严重事故的方法研究%Approach for Simulating Severe Accident of UO2-Zr Plate by SCDAP/RELAP5

    Institute of Scientific and Technical Information of China (English)

    张卓华; 彭诗念; 黄善仿; 于俊崇

    2013-01-01

    SCDAP/RELAP5是一种常见的机理性严重事故分析程序,能够分析多种类型的堆芯构件.通过对比分析SCDAP/RELAP5程序模拟棒形燃料元件与板型燃料元件堆芯在严重事故下行为的分析模型,结合UO2-Zr板型状元件堆芯的特性,提出了运用并改进SCDAP/RELAP5程序模拟UO2-Zr板型元件堆芯在严重事故下行为的研究方案.对程序结构的分析结果表明,SCDAP/RELAP5程序部分结构和模型适用于对UO2-Zr板型元件进行基本的严重事故分析,但需要通过创建新部件、研究新模型,并与已有模型的重新组合搭配才能较为精准地模拟UO2-Zr板型元件严重事故的实际行为.%As a common mechanistic code for safety analysis of severe accident,SCDAP/RELAP5 can simulate many types of core components phenomenon during severe accidents.Comparison of simulation model of fuel behavior under severe accident between fuel rod and ATR plate is described in this paper and the approach for simulating severe accident of UO2-Zr plate is concluded by combining structure properties of UO2-Zr.It is concluded that the basic analysis of severe accident of UO2-Zr plate could be achieved by S/R code from the code simulation.However,new core structure,new model of fuel behavior and combination of existing model should be developed in S/R code to simulate the precise core behavior of reactor assembled with UO2-Zr plate under severe accidents.

  13. RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

    Energy Technology Data Exchange (ETDEWEB)

    Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S. [B& W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

  14. New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

    Directory of Open Access Journals (Sweden)

    P. Balestra

    2016-01-01

    Full Text Available The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential, specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead, were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

  15. RELAP5/MOD3.2 post-test analysis and CIAU uncertainty evaluation of LOFT experiment L2-5

    International Nuclear Information System (INIS)

    Full text of publication follows: The paper deals with the activity performed at University of Pisa in the framework of the participation to the Phase II and III of the BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme. The scope of the Programme is to perform Large Break Loss-Of-Coolant Accident (LBLOCA) analyses making reference to experimental data and to a Nuclear Power Plant (NPP) in order to address the issue of 'the capabilities of computational tools' including scaling/uncertainty analysis. The justification for such an activity comes from the consideration that a wide spectrum of uncertainty methods applied to Best Estimate codes exist and are used in research laboratories, but their practicability and/or validity is not sufficiently established to support general use of the codes and acceptance by industry and safety authorities. The consideration of the Best Estimate codes and uncertainty evaluation for Design Basis Accident (DBA), by itself, shows the safety significance of the proposed activity.The Phase II of BEMUSE Programme is connected with the reanalysis of the Experiment L2 -5 performed in the LOFT (Loss Of Fluid Test) facility in June 1982. The LOFT facility, installed at the Idaho National Engineering Laboratory (INEL), is a 50 MWth Pressurized Water Reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT is typical of large (∼1000 MWe) commercial PWR. For the performance of Experiment L2-5, the LOFT facility was configured to simulate a double-ended 200 % cold leg break in a four-loop PWR operating at nominal conditions. Assumption of loss of offsite power and atypical primary coolant pump coast down were incorporated into the simulation to create core flow stagnation. The light water reactor transient analysis code Relap5/Mod3.2 has been used to simulate this experiment and the standard procedure adopted at

  16. RELAP5/MOD3.2 post-test analysis and CIAU uncertainty evaluation of LOFT experiment L2-5

    Energy Technology Data Exchange (ETDEWEB)

    Alessandro Petruzzi; Walter Giannotti [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2005-07-01

    Full text of publication follows: The paper deals with the activity performed at University of Pisa in the framework of the participation to the Phase II and III of the BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme. The scope of the Programme is to perform Large Break Loss-Of-Coolant Accident (LBLOCA) analyses making reference to experimental data and to a Nuclear Power Plant (NPP) in order to address the issue of 'the capabilities of computational tools' including scaling/uncertainty analysis. The justification for such an activity comes from the consideration that a wide spectrum of uncertainty methods applied to Best Estimate codes exist and are used in research laboratories, but their practicability and/or validity is not sufficiently established to support general use of the codes and acceptance by industry and safety authorities. The consideration of the Best Estimate codes and uncertainty evaluation for Design Basis Accident (DBA), by itself, shows the safety significance of the proposed activity.The Phase II of BEMUSE Programme is connected with the reanalysis of the Experiment L2 -5 performed in the LOFT (Loss Of Fluid Test) facility in June 1982. The LOFT facility, installed at the Idaho National Engineering Laboratory (INEL), is a 50 MWth Pressurized Water Reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT is typical of large ({approx}1000 MWe) commercial PWR. For the performance of Experiment L2-5, the LOFT facility was configured to simulate a double-ended 200 % cold leg break in a four-loop PWR operating at nominal conditions. Assumption of loss of offsite power and atypical primary coolant pump coast down were incorporated into the simulation to create core flow stagnation. The light water reactor transient analysis code Relap5/Mod3.2 has been used to simulate this experiment and the standard procedure

  17. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

  18. SCDAP/RELAP5 modeling of movement of melted material through porous debris in lower head

    International Nuclear Information System (INIS)

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head

  19. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 3. Application of RELAP5-3D and RELAP5/ MOD3.22 to Phase I of OECD PWR MSLB Benchmark

    International Nuclear Information System (INIS)

    same, except the initial steam generator mass inventory and the OTSG outlet temperature. The RELAP5-3D predicts a higher value of the mass in the steam generator, i.e., 28 094 kg, compared to the RELAP5/MOD3.22 calculation of 26 220 kg (Fig. 1). The reference value of the specifications is 28 395 kg. An important root cause of this difference is the different interfacial drag model used in the codes. In addition, a large initial fluid mass in the broken steam generator enhances the potential primary cooling system capability. The consequences of the calculation of the transient are mostly connected with the return-to-power (RTP) phenomenon (Fig. 2), where the total power versus time is represented. Although the first power peak has similar features for the two code runs, the second power peak predicted by RELAP5-3D is higher and shows a well-defined shape compared to the RELAP5/MOD3.22 calculation that is characterized by two peaks. Moreover, the core power after the second peak converges to a similar value at a similar time in both calculations. The other parameters that characterize the transient have similar times and tend mostly to coincide. The main conclusions of the analysis of the MSLB can be summarized as follows: 1. The adopted nuclear power plant (TMI-1) is safe enough in the considered scenario. 2. Both codes have predicted scram and RTP phenomena. The results obtained are qualitatively similar; however, in quantitative terms, noticeable differences have been found. In particular, 1. The initial steam generator mass and consequently the RTP peak are influenced by different thermal-hydraulic models governing the interfacial drag. 2. Concerning the different shape of the second power peak, one must notice that the beta version of RELAP5/MOD3.22 was adopted. Further calculations with the gamma version of RELAP5/MOD3.22, have shown that the second power peak is similar to that obtained with the RELAP5-3D code. 3. Because of the sensitive nature of this

  20. Modeling of control rod ejection transient for WWER-1000-model 446 using RELAP5m3.3/PARCSv2.6 coupled codes

    International Nuclear Information System (INIS)

    Highlights: • Capability to perform 3D neutronics/thermal–hydraulic analysis for WWER-1000 m446. • Good agreement was observed between the coupled codes results and FSAR data. • Capability to perform multi-dimensional analysis of complex transients such as a CREA. • WWER-1000 m446 shows a safe response during these transients performance. - Abstract: By using the Best Estimate (BE) method instead of conservative assumptions for the evaluation of reactor safety, significant economic considerations with optimal fuel burn-up could be obtained in addition to reactor safety. In this method, due to the detailed simulation and feedback considerations, special attention has been paid to the coupling of neutronic and thermo-hydraulic codes to achieve more reliable results. In this study, the Control Rod Ejection (CRE) transient has been simulated for Bushehr Nuclear Power Plant (BNPP) as a WWER-1000 power plant model 446 according to Final Safety Analysis Report (FSAR). CRE is a transient of Reactivity Initiated Accidents (RIA) category. In this study, the reactor thermo-hydraulic system has been simulated by RELAP5/mod3.3, while the neutron kinetic system of the reactor core has been simulated by the PARCSv2.6 code. These codes have been coupled utilizing Parallel Virtual Machine (PVM) interface software to consider the effects of thermal hydraulic and neutronic feedbacks. Thus, the power calculated by the PARCS code is used by the RELAP5 code and the obtained thermal hydraulic parameters are inserted to the PARCS code for macroscopic cross-section calculations. A computer program written by C++ has been used for the cycle execution of the WIMS code to produce the macroscopic cross-section library with the format required by the PARCS code. After the three-dimensional (3D) thermo-neutronic modeling of the reactor core, the Hot Zero Power (HZP) and Hot Full Power (HFP) versions of CRE transients, which have been considered in the plant’s FSAR, have been

  1. 基于RELAP5的沸水堆机理模型建立与仿真%Modeling and Simulation of BWR Mechanism Based on RELAP5

    Institute of Scientific and Technical Information of China (English)

    左琴; 陈彬; 邵之江; 周立芳; 钱积新

    2012-01-01

    核电厂高精度、实时、动态仿真要求有准确的反应堆热工水力以及控制系统等模型,它除用于电厂安全分析、操纵员培训等,还可用于控制系统参数优化、实际仪表控制系统验证等方面.为了研究在核电站控制系统问题上引入智能控制和优化算法的可行性,利用热工水力软件RELAP5,以沸水堆核电站为例,对沸水堆的堆芯系统和主给水系统进行了建模,并在稳态和暂态工况下进行了仿真验证.仿真结果与电站的实际数据基本一致,表明了基于RELAP5程序所建立的沸水堆热工水力模型的正确性.%The high - precision, real - time, dynamic simulation of a nuclear power plant needs accurate reactor thermal hydraulics and control system models, which can he used in safety analysis of power plant, operator training, control system parameter optimization, verification of the actual control instrument and so on. This paper studies the feasibility of introducing intelligent control methods and optimization algorithm to the control of nuclear power plant The model of Boiling Water Reactor nuclear power plant is built with the thermal hydraulic software RELAP5, including the reactor coolant system and the main feed water system. Then, simulations in steady and transient state are performed. The simulation results are consistent with the data collected from real plant, which validates the thermal - hydraulic system of Boiling Water Reactor based on RELAP5 developed in this paper.

  2. Implementation of Molten Salt Properties into RELAP5-3D/ATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Cliff Davis

    2005-01-01

    Molten salts are being considered as coolants for the Next Generation Nuclear Plant (NGNP) in both the reactor and the heat transport loop between the reactor and the hydrogen production plant because of their superior thermophysical properties compared to helium. Because specific molten salts have not been selected for either application, four separate molten salts were implemented into the RELAP5-3D/ATHENA computer program as working fluids. The implemented salts were LiF-BeF2 in a molar mixture that is 66% LiF and 34% BeF2, respectively, NaBF4-NaF (92% and 8%), LiF-NaF-KF (11.5%, 46.5%, and 42%), and NaF-ZrF4 (50% and 50%). LiF-BeF2 is currently the first choice for the primary coolant for the Advanced High- Temperature Reactor, while NaF-ZrF4 is being considered as an alternate. NaBF4-NaF and LiFNaF- KF are being considered as possible coolants for the heat transport loop. The molten salts were implemented into ATHENA using a simplified equation of state based on data and correlations obtained from Oak Ridge National Laboratory. The simplified equation of state assumes that the liquid density is a function of temperature and pressure and that the liquid heat capacity is constant. The vapor is assumed to have the same composition as the liquid and is assumed to be a perfect gas. The implementation of the thermodynamic properties into ATHENA for LiF-BeF2 was verified by comparisons with results from a detailed equation of state that utilized a soft-sphere model. The comparisons between the simplified and soft-sphere models were in reasonable agreement for liquid. The agreement for vapor properties was not nearly as good as that obtained for liquid. Large uncertainties are possible in the vapor properties because of a lack of experimental data. The simplified model used here is not expected to be accurate for boiling or single-phase vapor conditions. Because neither condition is expected during NGNP applications, the simplified equation of state is considered

  3. Simulation of a simple RCCS experiment with RELAP5-3D system code and computational fluid dynamics computer program

    International Nuclear Information System (INIS)

    A small scale experimental facility was designed to study the thermal hydraulic phenomena in the Reactor Cavity Cooling System (RCCS). The facility was scaled down from the full scale RCCS system by applying scaling laws. A set of RELAP5-3D simulations were performed to confirm the scaling calculations, and to refine and optimize the facility's configuration, instrumentation selection, and layout. Computational Fluid Dynamics (CFD) calculations using StarCCM+ were performed in order to study the flow patterns and two-phase water behavior in selected locations of the facility where expected complex flow structure occurs. (author)

  4. Analysis of integral circulation and decay heat removal experiments in the lead-bismuth CIRCE facility with RELAP5 code

    International Nuclear Information System (INIS)

    In this paper, the results of the post-test analysis of some integral circulation experiments conducted on the lead-bismuth CIRCE facility are presented in comparison with the experimental data. These experiments include the simulation of unprotected loss of flow and unprotected loss of heat sink transients in a pool-type heavy liquid metal reactor. Furthermore, the results of the pre-test analysis of a protected loss of heat sink and flow transient with decay heat removal by a heat exchanger immersed in the pool and operating in natural circulation is presented. All transient analyses have been performed with the RELAP5 thermal-hydraulic code. (author)

  5. Developmental assessment of RELAP5/MOD3.1 with separate-effect and integral test experiments: model changes and options

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.T. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-04-01

    A summary of modifications and options introduced in RELAP5/MOD3.1 (R5M3.1) is presented and it is shown that the predicting capabilities of the modified version of the code are greatly improved, while the general philosophy we followed in arriving at these modifications is also outlined. These changes which are the same ones we implemented in the past in the version 7j of the code, include 2 different heat transfer packages (one of them activated during reflooding), modification of the low mass-flux Groeneveld CHF look-up table and of the dispersed flow interfacial area (and shear) as well as of the criterion for transition into and out from this regime, almost complete elimination of the under-relaxation schemes of the interfacial closure coefficients etc. The modified R5M3.1 code is assessed against a number of separate-effect and integral test experiments and in contrast to the frozen version, is shown to result in physically sound predictions which are much closer to the measurements, while almost all the predicted variables are free of unphysical spurious oscillations. The modifications introduced solve a number of problems associated with the frozen version of the code and result in a version which can be confidently used both for SB-LOCA and LB-LOCA analyses. (author) 7 figs., 20 refs.

  6. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  7. Human cultured cells are capable to incorporate isolated plant mitochondria loaded with exogenous DNA

    Directory of Open Access Journals (Sweden)

    Laktionov P. P.

    2012-07-01

    Full Text Available Aim. To investigate the possibility of human cultured cells to incorporate isolated mitochondria together with exogenous DNA introduced into organelles. Methods. Two approaches were used for this purpose, fluorescent labelling of mitochondria and/or DNA with subsequent analysis of the cells subjected to incubation by microscopy or by quantitative PCR. Results. We have shown that human cultured cells lines, HeLa and HUVEC, are capable to uptake isolated plant mitochondria and that this process depends on the incubation time and concentration of organelles present in medium. The incorporated mitochondria can serve as vehicles to deliver exogenous DNA into human cells, this DNA is then distributed in different cell compartments. Conclusions. These results are preliminary and need further investigations, including testing the possibility of human cells to incorporate the mitochondria of human or animal origin and creating genetic construction which could provide certain selectivity or stability of the transferred exogenous DNA upon cell uptake of the mitochondria as vectors.

  8. Application of the coupled code RELAP5-QUABOX/CUBBOX in the system analysis of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bencik, V.; Feretic, D.; Debrecin, N. [Faculty of Electrical Engineering and Computing, Zagreb (Croatia)

    2002-11-01

    Best estimate codes and methods for the realistic simulation of operational transients and accidents are being developed in two directions. First, computer codes with models of the interaction between multidimensional neutron kinetic and NPP dynamic behavior enable realistic simulation of transients characterized by strong coupling between neutronics and thermal-hydraulics as well as of transients that result in asymmetrical spatial core power distribution. Coupled codes consisting of a system thermal-hydraulic code and a multidimensional neutronic code are being developed worldwide in order to accomplish that task. Secondly, development of the qualified plant nodalization and of the models of plant protection and control systems is important for the realistic system analysis of operational transients and accidents. Comparison of the coupled code and point kinetic results is important for the validation of the coupled code and to gain more experience in the use of the coupled code in realistic analyses. In this paper the results of two transients for NPP Krsko using the coupled code RELAP5-QUABOX/CUBBOX (R5QC) and RELAP5 stand alone code are discussed. (orig.)

  9. Conversion of the thermal hydraulics components of Almaraz NPP model from RELAP5 into TRAC-M

    International Nuclear Information System (INIS)

    In the scope of a joint project between the Spanish Nuclear Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) on the USNRC state-of-the-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper refers to the translation of the following thermal hydraulic models: pressurizer, hot and cold legs (including the pumps and the injection systems), and steam generators. The comparison of the results obtained with both codes shows a good agreement. However, some difficulties have been found in the translation of some code components, like pipes, heat structures, pumps, branchs, valves and boundary conditions. In this paper, these translation problems and their solutions are described.(author)

  10. Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

    Directory of Open Access Journals (Sweden)

    Skrzypek Maciej

    2015-09-01

    Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

  11. RELAP5/MOD3 assessment for calculation of safety and relief valve discharge piping hydrodynamic loads. International agreement report

    Energy Technology Data Exchange (ETDEWEB)

    Stubbe, E.J.; VanHoenacker, L.; Otero, R. [TRACTEBEL, Brussels (Belgium)

    1994-02-01

    This report presents an assessment study for the use of the code RELAP 5/MOD3/5M5 in the calculation of transient hydrodynamic loads on safety and relief discharge pipes. Its predecessor, RELAP 5/MOD1, was found adequate for this kind of calculations by EPRI. The hydrodynamic loads are very important for the discharge piping design because of the fast opening of the valves and the presence of liquid in the upstream loop seals. The code results are compared to experimental load measurements performed at the Combustion Engineering Laboratory in Windsor (US). Those measurements were part of the PWR Valve Test Program undertaken by EPRI after the TMI-2 accident. This particular kind of transients challenges the applicability of the following code models: two-phase choked discharge; interphase drag in conditions with large density gradients; heat transfer to metallic structures in fast changing conditions; two-phase flow at abrupt expansions. The code applicability to this kind of transients is investigated. Some sensitivity analyses to different code and model options are performed. Finally, the suitability of the code and some modeling guidelines are discussed.

  12. Assessment of 12 CHF prediction methods, for an axially non-uniform heat flux distribution, with the RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Ferrouk, M. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria)], E-mail: m_ferrouk@yahoo.fr; Aissani, S. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria); D' Auria, F.; DelNevo, A.; Salah, A. Bousbia [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (Italy)

    2008-10-15

    The present article covers the evaluation of the performance of twelve critical heat flux methods/correlations published in the open literature. The study concerns the simulation of an axially non-uniform heat flux distribution with the RELAP5 computer code in a single boiling water reactor channel benchmark problem. The nodalization scheme employed for the considered particular geometry, as modelled in RELAP5 code, is described. For this purpose a review of critical heat flux models/correlations applicable to non-uniform axial heat profile is provided. Simulation results using the RELAP5 code and those obtained from our computer program, based on three type predictions methods such as local conditions, F-factor and boiling length average approaches were compared.

  13. Assessment of 12 CHF prediction methods, for an axially non-uniform heat flux distribution, with the RELAP5 computer code

    International Nuclear Information System (INIS)

    The present article covers the evaluation of the performance of twelve critical heat flux methods/correlations published in the open literature. The study concerns the simulation of an axially non-uniform heat flux distribution with the RELAP5 computer code in a single boiling water reactor channel benchmark problem. The nodalization scheme employed for the considered particular geometry, as modelled in RELAP5 code, is described. For this purpose a review of critical heat flux models/correlations applicable to non-uniform axial heat profile is provided. Simulation results using the RELAP5 code and those obtained from our computer program, based on three type predictions methods such as local conditions, F-factor and boiling length average approaches were compared

  14. Simulation of the postulated stopping accident of the bombs of the primary circuit of Angra 2 with the code RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    This work presents the simulation of an anticipated transient for Angra 2 Nuclear Power Plant, where the coast down of the four reactor coolant pumps is verified. The best estimate thermal hydraulic system code RELAP5/MOD3.2 was used on this frame. A multi-purpose nodalization of Angra 2 was developed to simulate a comprehensive set of operational transients and accidents with RELAP5/MOD3.2 code. The overall objective of this work is to provide independent accident evaluation and further operational behavior follow-up to support the licensing process of the plant. (author)

  15. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 2. Sensitivity Studies for MSLB Exercises 2 and 3 with RELAP5/PANBOX

    International Nuclear Information System (INIS)

    As a contribution to the verification and validation of the RELAP5/PANBOX coupled code system (R/P/C), we took part in the Main-Steam-Line-Break (MSLB) Benchmark issued by OECD/NEA. Sensitivity studies with respect to external/ internal integration and coarse/fine channel representation have already been presented for exercise 2. The purpose of this paper is to extend the sensitivity studies to exercise 3 also and to present local results for safety-related parameters. R/P/C is a nuclear plant safety analysis code system that consists of the PANBOX core simulator coupled to the RELAP5 best-estimate plant simulator. The coupling is performed via the EUMOD RELAP5 interface package. R/P/C has the capabilities of RELAP5 with added ability for calculation of three-dimensional (3-D) neutronics and thermal margins with COBRA, the core thermal-hydraulic module of PANBOX. The neutronics nodalization is radially based on one node per fuel assembly (FA). Axially, 28 layers are modeled, where the specified mesh sizes are used with the exception of the 2 layers of 29.76 cm, which are subdivided into 4 layers. All calculations use the semi-analytical Nodal Expansion Method. The time discretization is based on the implicit Euler method combined with the exponential transformation technique. In the external integration of R/P/C, the core thermal-hydraulics solution is calculated by COBRA using core inlet boundary conditions from RELAP5. The channel geometry is based on one channel/FA, with axially 24 layers. In the internal integration, the core thermal-hydraulics solution is calculated by RELAP5. The channel geometry is based on 19 coarse channels with axially 11 core layers. R/P/C allows hot subchannel analysis by application of an on-line refinement of channels (HOSCAM). Fuel assembly powers, hot pin powers, and powers of a surrounding subchannel region are passed to COBRA for selected FAs in the external integration option. COBRA performs subchannel analysis by using a

  16. Study on a data source for fault diagnosis of nuclear power plant based on RELAP5%基于RELAP5的核动力装置故障诊断数据源研究

    Institute of Scientific and Technical Information of China (English)

    沈季; 夏虹; 苏应斌

    2009-01-01

    为解决故障状态下的核动力装置数据源问题,本文建立了核动力装置一、二回路系统的模型,选择秦山一期核电站为对象,利用RELAP5对蒸汽发生器U型管破裂进行计算.通过结果分析可知所建立的模型节点划分是合理的、数据卡编制准确,基于该模型产生的数据可信.将开发的数据与基于神经网络的故障诊断系统联调,诊断测试结果表明数据准确、充分,可以为核动力装置的故障诊断系统的研究提供数据支持.%The model of nuclear power plant primary and secondary circuits was established, taking Qinshan Unit 1 as the object. The fault of SG U-tube rupture was calculated based on RELAP5 for solving the fault data source problem. Through analyzing the calculation results, we can know that simulation node is reasonable, input data card is exact, and data based on this model is credible. The designed data combining with fault diagnosis system has been debugged. The result indicates that the data is exact and enough and can be one of the databases for study on fault diagnosis of nuclear power plant.

  17. SCDAP/RELAP5/MOD 3.1 code manual: User's guide and input manual. Volume 3

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume provides guidelines to code users based upon lessons learned during the developmental assessment process. A description of problem control and the installation process is included. Appendix a contains the description of the input requirements

  18. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    International Nuclear Information System (INIS)

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3DC/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  19. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide

  20. Simulation of a channel blockage transient in the Angra 2 Nuclear Reactor using a RELAP5-3D model

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Mantecon, Javier; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Pereira, Claubia; Reis, Patricia A.L.; Scari, Maria E., E-mail: mantecon1987@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: patricialire@yahoo.com.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The Angra 2 Nuclear Power Plant (NPP) is a Pressurized Water Reactor (PWR) type with electrical output of about 1350 MW. The RELAP5-3D code was used to develop a detailed thermal hydraulic model of such reactor using reference data from the Angra 2 Final Safety Analysis Report (FSAR). In this work, a blockage transient has been investigated at full power operation. The transient herein considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  1. Modeling a Helical-coil Steam Generator in RELAP5-3D for the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Nathan V. Hoffer; Piyush Sabharwall; Nolan A. Anderson

    2011-01-01

    Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary and secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.

  2. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    International Nuclear Information System (INIS)

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

  3. Assessment study of RELAP5/MOD2, CYCLE 36. 04 based on spray start-up test for DOEL-4

    Energy Technology Data Exchange (ETDEWEB)

    Moeyaert, P.; Stubbe, E.

    1989-07-01

    This report presents an assessment study for the code RELAP-5 MOD-2 based on a pressurizer spray start-up test of the Doel-4 power plant. Doel-4 is a three loop WESTINGHOUSE PWR plant ordered by the EBES utility with a nominal power rating of 1000 MWe and equipped with preheater type E steam generators. A large series of commissioning tests are normally performed on new plants, of which the so called pressurizer spray and heater test (SU-PR-01) was performed on February 2nd 1985. TRACTEBEL, being the Architect-Engineer for this plant was closely involved with all start-up tests and was responsible for the final approval of the tests.

  4. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  5. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras [Lithuanian Energy Institute, Kaunas (Lithuania)

    2003-07-01

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

  7. RELAP5 code study of ROSA/LSTF validation tests for PWR safety system using SG secondary-side depressurization

    International Nuclear Information System (INIS)

    RELAP5 code post-test analyses were performed on two ROSA/large scale test facility (LSTF) validation tests for PWR safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using steam generator (SG) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs a little after a safety injection signal. Auxiliary feedwater injection was done into the secondary-side of both SGs thereafter. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by boil-off. Core collapsed liquid level recovered after the initiation of accumulator (ACC) coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection (LPI) system. In the 4-in. break test, on the other hand, no core uncovery and heatup happened due to the coolant injection from the ACC and LPI systems. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code predicted well the overall trend of the major thermal-hydraulic response observed in the two LSTF tests. The code, however, overpredicted the peak cladding temperature (PCT) because of underprediction of the core collapsed liquid level due to inadequate prediction of the ACC flow rate in the 8-in. break case. Sensitivity analyses with the RELAP5 code indicated that a time delay for the SG depressurization start and break discharge coefficient for two-phase discharge flow affect the PCT significantly in the 8-in. break case. (author)

  8. Validation of the RELAP5 code for the modeling of flashing-induced instabilities under natural-circulation conditions using experimental data from the CIRCUS test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Helmholtz-Zentrum Dresden-Rossendorf e.V. (FZD), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany); Institute of Physics and Power Engineering, Obninsk (Russian Federation); Rohde, U., E-mail: U.Rohde@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf e.V. (FZD), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany); Manera, A. [Paul Scherrer Institute (Switzerland)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer We report about the simulation of flashing-induced instabilities in natural circulation systems. Black-Right-Pointing-Pointer Flashing-induced instabilities are of relevance for operation of pool-type reactors of small power at low pressure. Black-Right-Pointing-Pointer The RELAP5 code is validated against measurement data from natural circulation experiments. Black-Right-Pointing-Pointer The magnitude and frequency of the oscillations were reproduced in good agreement with the measurement data. - Abstract: This paper reports on the use of the RELAP5 code for the simulation of flashing-induced instabilities in natural circulation systems. The RELAP 5 code is intended to be used for the simulation of transient processes in the Russian RUTA reactor concept operating at atmospheric pressure with forced convection of coolant. However, during transient processes, natural circulation with flashing-induced instabilities might occur. The RELAP5 code is validated against measurement data from natural circulation experiments performed within the framework of a European project (NACUSP) on the CIRCUS facility. The facility, built at the Delft University of Technology in The Netherlands, is a water/steam 1:1 height-scaled loop of a typical natural-circulation-cooled BWR. It was shown that the RELAP5 code is able to model all relevant phenomena related to flashing induced instabilities. The magnitude and frequency of the oscillations were reproduced in a good agreement with the measurement data. The close correspondence to the experiments was reached by detailed modeling of all components of the CIRCUS facility including the heat exchanger, the buffer vessel and the steam dome at the top of the facility.

  9. Total Transfer Capability Assessment Incorporating Corrective Controls for Transient Stability using TSCOPF

    Science.gov (United States)

    Hakim, Lukmanul; Kubokawa, Junji; Yorino, Naoto; Zoka, Yoshifumi; Sasaki, Yutaka

    Advancements have been made towards inclusion of both static and dynamic security into transfer capability calculation. However, to the authors' knowledge, work on considering corrective controls into the calculation has not been reported yet. Therefore, we propose a Total Transfer Capability (TTC) assessment considering transient stability corrective controls. The method is based on the Newton interior point method for nonlinear programming and transfer capability is approached as a maximization of power transfer with both static and transient stability constraints are incorporated into our Transient Stability Constrained Optimal Power Flow (TSCOPF) formulation. An interconnected power system is simulated to be subjected to a severe unbalanced 3-phase 4-line to ground fault and following the fault, generator and load are shed in a pre-defined sequence to mimic actual corrective controls. In a deregulated electricity market, both generator companies and large load customers are encouraged to actively participate in maintaining power system stability as corrective controls upon agreement of compensation for being shed following a disturbance. Implementation of this proposal on the actual power system operation should be carried out through combining it with the existing transient stabilization controller system. Utilization of these corrective controls results in increasing TTC as suggested in our numerical simulation. As Lagrange multipliers can also describe sensitivity of both inequality and equality constraints to the objective function, then selection of which generator or load to be shed can be carried out on the basis of values of Lagrange multipliers of its respective generator's rotor angle stability and active power balance equation. Hence, the proposal in this paper can be utilized by system operator to assess the maximum TTC for specific loads and network conditions.

  10. Safety Analysis on Dual-functional Lithium Lead Test Blanket Module With RELAP5%基于 RELAP5的双功能液态锂铅实验包层模块安全分析

    Institute of Scientific and Technical Information of China (English)

    李伟; 田文喜; 秋穗正; 苏光辉

    2013-01-01

    利用嵌入了液态锂铅(LiPb)的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅(DFLL)实验包层模块(TBM)的安全特性进行评价。对DFLL-TBM 及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁(FW )结构材料表面最高温度低于允许值550℃。事故工况下氦气泄漏引起的ITER真空室(VV)、窗口设备室(port cell)以及托卡马克冷却水系统大厅拱顶(TCWS vault)的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM 的设计可满足ITER对其热工水力安全方面的要求。%Safety assessment on the dual-functional lithium lead test blanket module (DFLL-TBM) was performed with a modified version of RELAP5/MOD3 code in which the LiPb eutectic thermal-hydraulic sub-module was inserted .The DFLL-TBM and its ancillary cooling systems were modeled to conduct the computation and analysis for steady-state operation ,anticipated operational incidents and relevant accidents .Compu-tational results indicate that the maximum surface temperature of the first wall (FW) structural material is lower than the allowable value of 550 ℃ .For the accident analy-ses ,none of the pressure increases in ITER vacuum vessel (VV) ,port cell and TCWS vault induced by helium leaking is beyond the ITER safety limit of 0.2 MPa .No melting of the TBM box is found and the decay heat can be removed efficiently by the radiation heat transfer .With the current design ,DFLL-TBM can meet the thermal-hydraulic safety requirements from IT ER .

  11. Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Andrew M. Ward

    2011-01-01

    Full Text Available In order to analyze the steady state and transient behavior of CNA-II, several tasks were required. Methods and models were developed in several areas. HELIOS lattice models were developed and benchmarked against WIMS/MCNP5 results generated by NA-SA. Cross-sections for the coupled RELAP5/PARCS calculation were extracted from HELIOS within the GenPMAXS framework. The validation of both HELIOS and PARCS was performed primarily by comparisons to WIMS/PUMA and MCNP for idealized models. Special methods were developed to model the control rods and boron injection systems of CNA-II. The insertion of the rods is oblique, and a special routine was added to PARCS to treat this effect. CFD results combined with specialized mapping routines were used to model the boron injection system. In all cases there was good agreement in the results which provided confidence in the neutronics methods and modeling. A coupled code benchmark between U of M and U of Pisa is ongoing and results are still preliminary. Under a LOCA transient, the best estimate behavior of the core appears to be acceptable.

  12. PSA support safety analysis using RELAP5 for the reactivity insertion event in 14 MW TRIGA reactor

    International Nuclear Information System (INIS)

    The paper presents the deterministic support analysis in case of Reactivity Insertion Accident (RIA) considered as initiating event in the PSA project. It studies the reactivity worth necessary to damage the research reactor fuel. In a previous PSA study the postulated initiating event due to a mistake in fuel handling was assumed as resulting from falling of one or at most two fuel bundles from the lifting device operated for core configuration rearrangements. This type of event was actually oc curing in the nineties. The paper gives some elements of the previous PSA model with respect to this initiating event, the event tree and the results of the accident aftermath. The focus of the paper is on the results of applying the thermalhydraulic code RELAP5 Mod 3.2 which uses a point kinetics model for studying the transient in case of different reactivity worths and insertion times. The results include evolutions of heat transfer mode, maximum temperature inside fuel elements and peak values of the power excursion. The conclusions highlight the possibility of infringement of the safety criteria for the TRIGA Ssr 14 MW reactor during the analyzed transients and also discuss the necessity of including this event in the PSA model. (authors)

  13. Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

    International Nuclear Information System (INIS)

    This report summarizes significant technical findings from the LP-FP-2 Experiment sponsored by OECD. It was the second, and final, fission product experiment conducted in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory. The overall technical objective of the test was to contribute to the understanding of fuel rod behavior, hydrogen generation, and fission product release, transport, and deposition during a V-sequence accident scenario that resulted in severe core damage. An 11 by 11 test bundle, comprised of 100 pre-pressurized fuel rods, 11 control rods, and 10 instrumented guide tubes, was surrounded by an insulating shroud and contained in a specially designed central fuel module, that was inserted into the LOFT reactor. The simulated transient was a V-sequence loss-of-coolant accident scenario featuring a pipe break in the low pressure injection system line attached to the hot leg of the LOFT broken loop piping. The transient was terminated by reflood of the reactor vessel when the outer wall shroud temperature reached 1517 K. With sustained fission power and heat from oxidation and metal-water reactions, elevated temperatures resulted in zircaloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. A description and evaluation of the major phenomena, based upon the response of on line instrumentation, analysis of fission product data, post-irradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented

  14. Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

    Science.gov (United States)

    Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

    2012-10-01

    Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

  15. Analysis of the IRIS pressurizer behavior in the presence of noncondensable gases using RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Eduarda da C.A.; Castrillo, Lazara S., E-mail: e.camedeiros@gmail.com, E-mail: lazara@poli.br [Universidade de Pernambuco, Recife, PE (Brazil). Escola Politecnica. Departamento de Engenharia Mecanica

    2015-07-01

    Insurge and outsurge phenomena are transient states and could be analyzed by thermodynamics principles, the pressurizer behavior will vary in response to mass flow changes. These surges can occur in the presence of noncondensable gases. On this paper, with the code RELAP5, the IRIS reactor pressurizer is described to analyze surges phenomena in their control volumes with non-condensable gases since they modify the pressure response. A set of three pipes components represents the pressurizer regions, connected with each other by single junctions components, the bottom volume control is connected to the primary circuit, represented by a time dependent volume component, through a time dependent junction component, which describes the mass flow behavior during surges through surge orifices. The hydrodynamic components representing the pressurizer are surrounded by heat structures, in addition there are heat structures inside the bottom volume control describing the behavior of electrical heaters, that operate in cases of outsurges. The analysis are intended to detail the behavior variables as pressure, temperature and volume of liquid inside the pressurizer during a water surge coming from the primary circuit or a water surge coming from the pressurizer to the primary circuit. (author)

  16. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  17. Implementation of a New DTSTEP Algorithm for use in RELAP5-3D and PVMEXEC Completion Report

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2010-12-01

    The PVM Coupling methodology for decomposing a complex model into domains onto which individual programs may be applied has proven effective for solving many multi-physics problems. There have been, from the outset, some detailed and/or long-running models that cause the process to fail. This project addressed the PVM coupling issues surrounding the DTSTEP subroutines on RELAP5-3D and PVMEXEC. Some 25 errors are listed in Tables 1 and 18 and in Section 11. These arise from deficiencies in the floating point calculation and testing of time steps, cumulative time, and time targets. The algorithmic replacement of floating point control of these items with integer based timestepping was developed and implemented. The result of the first phase, undertaken by the INL was that all but three of these issues were resolved. Moreover, two conceptual errors in DTSTEP that were not PVM coupling related were discovered and solved. The final, and most difficult three PVM Bettis User Problems, were solved during the Bettis phase of development and debugging. In 8 months since the conclusion of the project, no further DTSTEP related PVM Coupling errors have been reported.

  18. Analysis of Air-Water Two Phase Flow for K-HERMES-HALF Experiment using RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Ha, Kwang Soon; Kim, Sang Baik; Hong, Seong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Heo, Sun [KHNP Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The IVR (In-Vessel corium Retention) through the ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the integrity of the reactor pressure vessel during a severe accident in a nuclear power plant. This measure has been adopted in some low-power reactors such as the AP600, AP1000, and the Loviisa nuclear power plants as a design feature, and in the high-power reactor of the APR (Advanced Power Reactor) 1400 as an accident management strategy for severe accident mitigation. As part of a study on two-phase flow in the reactor cavity under external reactor vessel cooling in the APR1400, K-HERMES-HALF experiment (Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale) had performed at KAERI. This large-scale experiment using a half-height and half-sector model of the APR1400 uses the non-heating method of the air injection. In this research, K-HERMES-HALF test results had been evaluated by using RELAP5/MOD3 computer code to observe and evaluate the two-phase natural circulation phenomena through the annulus gap between the outer reactor vessel and the vessel insulation material

  19. Assessment study of RELAP5/MOD2 Cycle 36. 04 based on pressurizer safety and relief valve tests

    Energy Technology Data Exchange (ETDEWEB)

    Stubbe, E.J.; Vanhoenacker, L.

    1990-07-01

    This report presents a code assessment study based on full size relief and assisted safety valve (called SEBIM) tests performed on the CUMULUS valve test rig operated by Electricite de France (EdF). The increased awareness that the pressuriser safety and relief valves are not reliable under water blowdown conditions, has led to the design, testing and installation of so called assisted safety valves of which the SEBIM (TM) valves are an example. These valves, used in tandem, are gradually replacing the safety and relief valves on pressurisers in some European PWR's. Before installation at the plant, the Belgian safety authorities requested a thorough full scale testing of these valves on a test rig (CUMULUS) equipped with sufficient diagnostics to measure the characteristics of the valve. The Belgian architect-engineering firm TRACTEBEL was called upon the specify, order and test these valves for installation at the DOEL 1 and DOEL 2 power plants. These tests do provide sufficient data of high quality to justify an assessment study of the code RELAP-5 MOD-2 CYCLE 36 in the ICAP framework which is the subject of this report.

  20. RGUI 1.0, New Graphical User Interface for RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    G. L. Mesina; J. Galbraith

    1999-04-01

    With the advent of three-dimensional modeling in nuclear safety analysis codes, the need has arisen for a new display methodology. Currently, analysts either sort through voluminous numerical displays of data at points in a region, or view color coded interpretations of the data on a two-dimensional rendition of the plant. RGUI 1.0 provides 3D capability for displaying data. The 3D isometric hydrodynamic image is built automatically from the input deck without additional input from the user. Standard view change features allow the user to focus on only the important data. Familiar features that are standard to the nuclear industry, such as run, interact, and monitor, are included. RGUI 1.0 reduces the difficulty of analyzing complex three-dimensional plants.

  1. RGUI 1.0, New Graphical User Interface for RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Mesina, George Lee; Galbraith, James Andrew

    1999-04-01

    With the advent of three-dimensional modeling in nuclear safety analysis codes, the need has arisen for a new display methodology. Currently, analysts either sort through voluminous numerical displays of data at points in a region, or view color coded interpretations of the data on a two-dimensional rendition of the plant. RGUI 1.0 provides 3D capability for displaying data. The 3D isometric hydrodynamic image is built automatically from the input deck without additional input from the user. Standard view change features allow the user to focus on only the important data. Familiar features that are standard to the nuclear industry, such as run, interact, and monitor, are included. RGUI 1.0 reduces the difficulty of analyzing complex three dimensional plants.

  2. A comparison of the effect of the first and second upwind schemes on the predictions of the modified RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.Th. [Paul Scherrer Institute (PSI), Villigen (Switzerland)

    1995-09-01

    As is well-known, both TRAC-BF1 and TRAC-PF are using the first upwind scheme when finite-differencing the phasic momentum equations. In contrast, RELAP5 uses the second upwind which is less diffusive. In this work, we shall assess the differences between the two schemes with our modified version of RELAP5/MOD3 by analyzing some transients of interest. These will include the LOFT LP-LB-1 and LOBI small break LOCA (SB-LOCA) BL34 tests, and a commercial PWR 200% hypothetical large break LOCA (LB-LOCA). In particular, we shall show that for some of these transients, the employment of the first upwind scheme results in significantly different code predictions than the ones obtained when the second upwind scheme is used.

  3. Simulation of research loop LOBI-MOD2 with RELAP5/MOD3.3 code for LOBI thermo hydraulic test A1-93

    Energy Technology Data Exchange (ETDEWEB)

    Pesaran, Farshad; Barati, Ramin [Islamic Azad Univ., Shiraz (Iran, Islamic Republic of). Dept. of Electrical Engineering

    2016-06-15

    RELAP5/MOD3.3 is one of the used computer codes for the simulation of event thermal-hydraulics of nuclear power plants. The LOBI test facility is a full-power high-pressure integral system test facility, representing an approximately 1: 700 scale model of a 4-loop, 1300 MWe PWR. A new simulation of the small break LOCA test A1-93 has been carried out in a LOBI/Mod2 facility for reaching good agreement and to evaluate the performance of the RELAP5/MOD3.3 code. Good agreement was obtained in general between the code predictions and the experimental data in transient state.

  4. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  5. RELAP5-3D Developmental Assessment: Comparison of Versions 4.2.1i and 4.1.3i

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    2014-06-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.2.1i and 4.1.3i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  6. Comparison of an integral response scaling method with Ishii's scaling method and its validation using RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    An integral response scaling method for a reduced-height test facility is suggested and the scaling laws derived from it are compared with Ishii's scaling. In the present scaling method it turns out that flow velocities in the vertical channel and through the break area or injection area should be preserved. RELAP5/MOD3.2 code calculations of pot-boiling, blowdown, heat transfer in Steam Generator(SG) and off-take are conducted for the validation of the present scaling method. Four scaled-down models are designed based on the present method and Ishii's scaling method given length and area scales of 1/5 and 1/100, respectively. RELAP5/MOD3.2 calculations show that the scaled-down model based on the present scaling method well maintains the similarity of the nondimensional mixture level in pot-boiling, the nondimensional pressure in blowdown and the heat transfer coefficient in SG

  7. RELAP5-3D Developmental Assessment: Comparison of Versions 4.3.4i and 4.2.1i

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.3.4i and 4.2.1i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  8. Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

  9. Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

    OpenAIRE

    Siniša Šadek; Srđan Špalj; Bruno Glaser

    2010-01-01

    RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from ...

  10. Assessment of RELAP5/MOD3/V5M5 against the UPTF Test No. 11 (countercurrent flow in PWR hot leg)

    International Nuclear Information System (INIS)

    Analysis of the UPTF Test No. 11 using the open-quotes best-estimateclose quotes computer code RELAP5/MOD3/Version 5M5 is presented. Test No. 11 was a quasi-steady state, separate effect test designed to investigate the conditions for countercurrent flow of steam and saturated water in the hot leg of a PWR. Without using the code's new countercurrent flow limitation (CCFL) model, RELAP5/MOD3/V5M5 overestimated the mass flow rate of back down flowing water up to 35% (1.5 MPa runs) and 43% (0.3 MPa runs). This is the most obvious difference to RELAP5/MOD2, which did not allow enough countercurrent flow. From the point of view of performing plant calculations this is certainty an improvement, because the new junction-based CCFL option could be used to restrict the flows to a flooding curve defined by a user-supplied correlation. Very good agreement with the experimental data for 1.5 MPa -- which are relevant for SBLOCA reflux condensation conditions -- could be obtained using the code's new CCFL option in the middle of the inclined part (riser) of the hot leg. Using the same CCFL correlation for the simulation of 0.3 MPa test series -- typical for reflood conditions -- the code underestimated by 44% the steam mass flow rate at which complete liquid carry over occurs. An unphysical result was received using a CCFL correlation of the Wallis type with the intercept C = 0.644 and the slope m = 0.8. The unphysical prediction is an indication of possible programming errors in the CCFL model of the RELAP5/MOD3/V5M5 computer code

  11. Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

  12. An analysis of ROSA-IV/LSTF 10% main steam line break test run SB-SL-01 using RELAP5/MOD3

    International Nuclear Information System (INIS)

    This paper presents RELAP5/MOD3 code calculations of a 10% main steam line break test, designated as RUN SB-SL-01, conducted using the ROSA-4 Large Scale Test Facility (LSTF). The RELAP5/MOD3 input deck of LSTF, which includes 189 volumes, 200 junctions, and 180 heat slabs, was modeled to obtain best-estimate predictions of several important features during the main steam line break accident in order to property evaluate the consequences of this accident. The main conclusions drawn were that the results of RELAP5/MOD3 code calculations were in reasonable agreements with test RUN SB-SL-01, especially for the trends of key parameters. Detailed investigations indicated minor discrepancies in RCS pressure during the period of time that voiding occurred in the upper head. This is possible due to emptying of the pressurizer and voiding in the upper head. Sensitivity studies were also performed for the break junction discharge coefficient and the separator drain line loss coefficient. These parameters had significant effects on the steam quality on the secondary side and on the break flow through the change of water inventory on the secondary side. This phase separation process was adequately predicted during all transients with break junction discharge coefficient of 0.85 and separator drain line loss coefficient of 10

  13. The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

    Energy Technology Data Exchange (ETDEWEB)

    Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

  14. RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Pavlova, M.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlova@inrne.bas.bg; Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, A.E. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg; Gencheva, R.V. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: roseh@inrne.bas.bg

    2006-02-15

    This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit. RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters. This paper discusses the results of the thermal-hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences. This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP)

  15. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)

    2007-09-15

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

  16. Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) for ITER Using System Code RELAP5

    Institute of Scientific and Technical Information of China (English)

    Jin Xuezhou; R. Meyder

    2005-01-01

    The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.

  17. RELAP5/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

    International Nuclear Information System (INIS)

    RELAP5/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients. In developing the 4-loop plant model, a three loop W plant model was modified at ANL into a four loop plant. This plant model retained the two loop nodalization of the original model, but split them such that one loop represented the faulted steam generator (broken at the outlet just past the integral flow restrictor) while the other loop represented all three of the intact loops. In the case of the 3-loop plant model, a RELAP5 model was modified to perform this replication calculation. One loop represented the faulted steam generator (broken between the steam generator and the non-integral flow restrictor) while the other loop represented the remaining two intact loops. In both cases (3-loop and 4-loop), the reactor vessel was modelled as two parallel channels to accommodate the Westinghouse steam line break methodology

  18. Analysis of BWR instabilities coupled with 3D code RELAP5 / PARCSv2.7. Application to the event happened in Oskarshamn-2 in 1999; Analisis de inestabilidades en BWR con el codigo acoplado 3D RELAP5/PARCSv2.7. Aplicacion al evento sucedido en Oskarshamn-2 en1999

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fenoll, M.; Barrachina, T.; Miro, R.; Verdu, G.

    2014-07-01

    In this work, part of our works in the frame of the OECD/NEA Oskarshamn-2 (O{sub 2}) BWR Stability Benchmark for Coupled Code Calculations and Uncertainty Analysis in Modelling are shown. The objective is to simulate the instability event registered in February 1999 at the Swedish NPP Oskarshamn-2 with the coupled code RELAP5/PARCSv2.7. (Author)

  19. Validation of One-Dimensional Module of MARS-KS1.2 Computer Code By Comparison with the RELAP5/MOD3.3/patch3 Developmental Assessment Results

    International Nuclear Information System (INIS)

    This report records the results of the code validation for the one-dimensional module of the MARS-KS thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 Code Developmental Assessment Problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

  20. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E.J.; Chung, B.D.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1993-06-01

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator.

  1. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    International Nuclear Information System (INIS)

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator

  2. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

    International Nuclear Information System (INIS)

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  3. RELAP5/MOD3.3程序对非能动核电厂小破口失水事故的适用性研究%Applicability Research of RELAP5/MOD3.3 for Small Break Loss of Coolant Accident of NPP With Passive Safety System

    Institute of Scientific and Technical Information of China (English)

    徐财红; 史国宝

    2014-01-01

    AP1000核电厂采用非能动堆芯冷却系统来缓解小破口失水事故(SBLOCA),缓解事故的理念是流动冷却。RELAP5/MOD3.3程序适用于传统核电厂SBLOCA 研究,对于非能动电厂SBLOCA研究的适用性需重新研究与评估。本工作基于非能动电厂小破口失水事故的分析,结合RELAP5/MOD3.3的结构与模型,对其进行评估和改进。为验证改进后的REL A P5/M OD3.3的适用性,以A P1000小破口失水事故的验证试验台架APEX-1000为模拟对象,分析模拟DBA-02、NRC-05事故工况。分析结果表明,改进后的REL A P5/M OD3.3的计算结果与试验数据符合较好。%The passive core cooling system is used in AP 1000 to mitigate the small break loss of coolant accident (SBLOCA) .The RELAP5/MOD3.3 code is generally applicable to the traditional NPP SBLOCA research , but for the passive NPP SBLOCA , its applicability will need further study and evaluation . Based on the analysis of the important phenomenon of the SBLOCA of the passive NPP , the RELAP5/MOD3.3 code was assessed and modified . In order to verify the applicability of the modified RELAP5/MOD3.3 code ,the DBA-02 and NRC-05 cases of APEX-1000 which was the test facility for verifying AP1000 small break loss of coolant accident ,were simulated . It shows good agreement between the results of the modified RELAP5/MOD3.3 code and experiment data .

  4. MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code

    International Nuclear Information System (INIS)

    For safety analyses to support conversion of MNSR reactors from HEU fuel to LEU fuel, a RELAP5-3D model was set up to simulate the entire MNSR system. This model includes the core, the beryllium reflectors, the water in the tank and the water in the surrounding pool. The MCNP code was used to obtain the power distributions in the core and to obtain reactivity feedback coefficients for the transient analyses. The RELAP5-3D model was validated by comparing measured and calculated data for the NIRR-1 reactor in Nigeria. Comparisons include normal operation at constant power and a 3.77 mk rod withdrawal transient. Excellent agreement was obtained for core coolant inlet and outlet temperatures for operation at constant power, and for power level, coolant inlet temperature, and coolant outlet temperature for the rod withdrawal transient. In addition to the negative reactivity feedbacks from increasing core moderator and fuel temperatures, it was necessary to calculate and include positive reactivity feedback from temperature changes in the radial beryllium reflector and changes in the temperature and density of the water in the tank above the core and at the side of the core. The validated RELAP5-3D model was then used to analyze 3.77 mk rod withdrawal transients for LEU cores with two UO2 fuel pin designs. The impact of cracking of oxide LEU fuel is discussed. In addition, steady-state power operation at elevated power levels was evaluated to determine steady-state safety margins for onset of nucleate boiling and for onset of significant voiding. (author)

  5. Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P. [Asociacion Nuclear Asco, Barcelona (Spain)

    1993-12-01

    The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ``Pressurizer spray valve faulty opening`` presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data.

  6. Restructuring the electronic medical record to incorporate full digital signature capability.

    OpenAIRE

    Zuckerman, A. E.

    2001-01-01

    The security of Electronic Medical Records can be enhanced by the addition of digital signatures that guarantee data integrity, authenticate the signer, and establish non-repudiation through the use of public key encryption. The task is complicated by the contribution of multiple providers to an encounter and the entry of data at multiple points in time Dividing encounters into an episode of care and redesigning the data model of the EMR will facilitate full signature capabilities. Generation...

  7. Incorporation of Multi-Member Substructure Capabilities in FAST for Analysis of Offshore Wind Turbines: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Song, H.; Robertson, A.; Jonkman, J.; Sewell, D.

    2012-05-01

    FAST, developed by the National Renewable Energy Laboratory (NREL), is an aero-hydro-servo-elastic tool widely used for analyzing onshore and offshore wind turbines. This paper discusses recent modifications made to FAST to enable the examination of offshore wind turbines with fixed-bottom, multi-member support structures (which are commonly used in transitional-depth waters).; This paper addresses the methods used for incorporating the hydrostatic and hydrodynamic loading on multi-member structures in FAST through its hydronamic loading module, HydroDyn. Modeling of the hydrodynamic loads was accomplished through the incorporation of Morison and buoyancy loads on the support structures. Issues addressed include how to model loads at the joints of intersecting members and on tapered and tilted members of the support structure. Three example structures are modeled to test and verify the solutions generated by the modifications to HydroDyn, including a monopile, tripod, and jacket structure. Verification is achieved through comparison of the results to a computational fluid dynamics (CFD)-derived solution using the commercial software tool STAR-CCM+.

  8. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

  9. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.T. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  10. RELAP5-3D Modeling of Heat Transfer Components (Intermediate Heat Exchanger and Helical-Coil Steam Generator) for NGNP Application

    International Nuclear Information System (INIS)

    The Next Generation Nuclear Plant project is aimed at the research and development of a helium-cooled high-temperature gas reactor that could generate both electricity and process heat for the production of hydrogen. The heat from the high-temperature primary loop must be transferred via an intermediate heat exchanger to a secondary loop. Using RELAP5-3D, a model was developed for two of the heat exchanger options a printed-circuit heat exchanger and a helical-coil steam generator. The RELAP5-3D models were used to simulate an exponential decrease in pressure over a 20 second period. The results of this loss of coolant analysis indicate that heat is initially transferred from the primary loop to the secondary loop, but after the decrease in pressure in the primary loop the heat is transferred from the secondary loop to the primary loop. A high-temperature gas reactor model should be developed and connected to the heat transfer component to simulate other transients

  11. Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

    International Nuclear Information System (INIS)

    As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

  12. Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

    Directory of Open Access Journals (Sweden)

    Siniša Šadek

    2010-01-01

    Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

  13. Conversion of control systems, protection and engineering safeguard system signals of Almaraz NPP model from RELAP5 into TRAC-M

    International Nuclear Information System (INIS)

    In the scope of a joint project between the Spanish Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) about the USNRC state-of-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As a part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. One of the goals of the project is to analyze the conversion of control blocks, signal variables and trips in order to correct modelling all instrumentation and control systems, and also protection and engineering safeguard system-signals of the NPP. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper describes the problems found in the conversion and the solutions achieved.(author)

  14. Vectorization and improvement of nuclear codes (MEUDAS4, FORCE, STREAM V2.6, HEATING7-VP, SCDAP/RELAP5/MOD2.5, NBI3DGFN)

    International Nuclear Information System (INIS)

    Eight nuclear codes have been vectorized and modified to improve their performance. These codes are magnetic fluid equilibrium code MEUDAS4 (CR and FFT versions), the magnetic field analysis code FORCE, the three-dimensional heat fluid analysis code STREAM V2.6, the three-dimensional heat analysis code HEATING 7-VP, the severe accident transient analysis code SCDAP/RELAP 5/MOD 2.5 for light water reactors, the ion beam orbital analysis code NBI3DGFN, and a free electron laser analysis code. The speedup ratios of the vectorized versions to the original ones in scalar mode are 2.3-4.9, 1.9-5.4, 2.6-6.2, and 1.9 for the MEUDAS4, STREAM, FORCE, and free electron laser analysis code, respectively. The definition method of the computational regions in the HEATING7-VP is improved. The SCDAP/RELAP5/MOD2.5 is modified to use extended memory regions of the computer. In this report, outlines of the codes, techniques used in the vectorization and reorganization of the codes, verification of computed results, and improvement on the performance are presented. (author)

  15. Simulation with RELAP5/MOD3.3 of a postulated 10% hot leg break in Angra 2 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: cvga@cdtn.br, e-mail: etp@cdtn.br, e-mail: aroneid@cdtn.br

    2009-07-01

    This paper presents the simulation results of a 10% break in the hot leg of Angra 2 nuclear power plant, which was run with the computer code RELAP5/MOD3.3. The initial steady state conditions for this simulation are in agreement with the experiment named SB-HL-02 that was conducted in the Large Scale Test Facility in the Rig of Safety Assessment-IV program (ROSA-IV/LSTF). The main boundary conditions specified for the simulation were: high pressure injection system (HPI) and auxiliary feedwater system (AFW) were assumed to be unavailable; and loss of offsite power was assumed to occur concurrently with scram. The results obtained were scaled down and compared with the ROSA-IV/LSTF test, which was performed with the same boundary conditions. This activity was executed in the scope of IAEA research project (CRP J72005) - Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3.3 Code Applying CIAU Methodology. (author)

  16. RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2012-01-01

    Full Text Available A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG secondary-side pressures were maintained, respectively, at around 16 and 8 MPa by cycle opening of pressurizer (PZR power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

  17. Restructuring the electronic medical record to incorporate full digital signature capability.

    Science.gov (United States)

    Zuckerman, A E

    2001-01-01

    The security of Electronic Medical Records can be enhanced by the addition of digital signatures that guarantee data integrity, authenticate the signer, and establish non-repudiation through the use of public key encryption. The task is complicated by the contribution of multiple providers to an encounter and the entry of data at multiple points in time Dividing encounters into an episode of care and redesigning the data model of the EMR will facilitate full signature capabilities. Generation of digital signatures is best accomplished using microprocessors on smart cards that control visibility of the private keys and assist in user authentication. The Java Programming Language including cryptography extensions and a smart card API is a useful tool for adding digital signature to an EMR. Inter-operability of signatures and continuity of signature will require attention to standards and preservation of cryptography and authentication certificate archives. Digital signatures will need to accommodate changes in data storage formats when information is transported between EMR systems using XML or other transaction standards because the original signatures will not validate if the data storage format changes. The costs of adding digital signature to EMR mandates serious examination of the business case for digital signature within an EMR as compared with transactions such as electronic prescriptions. At present, there is no regulatory requirement for digital signature of an EMR. PMID:11825294

  18. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty; Qualificacao e aplicacao de codigo de acidentes de reatores nucleares com capacidade interna de avaliacao de incerteza

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Ronaldo Celem

    2001-10-15

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  19. The addition of non-condensable gases into RELAP5-3D for analysis of high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of- coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide

  20. The application of Cathare 1 V1.3 to LOBI small break Loca experiments and a comparison with RELAP5/MOD2

    International Nuclear Information System (INIS)

    The paper presents an overview of the application of CATHARE V1.3 to LOBI Small Break LOCA tests, performed at Dipartimento di Costruzioni Meccaniche e Nucleari of Pisa University. In particular, the development of a new nodalization of LOBI facility is discussed along with the analysis of tests A2-81 (1% CL break). A1-83 (10% CL break) and A1-84 (10% HL break). In the second part of the paper, uncertainties are outlined which are typical of the analysis of experiments in integral test facilities. Finally, on the basis of the application of RELAP5/MOD2 to the analysis of test A2-81, a judgement is given about the behaviour of the two codes emphasizing the related advantages and disadvantages

  1. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  2. Simulation of steam condensation in the presence of noncondensable gases in horizontal condenser tubes using RELAP5 for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Horizontal heat exchangers are used in advanced light water nuclear reactors in their passive cooling systems, such as residual heat removal (RHRS) and passive containment cooling system (PCCS). Condensation studies of steam and noncondensable gases mixtures in these heat exchangers are very important due to the phenomena multidimensional nature and the condensate stratification effects. This work presents a comparison between simulation results and experimental data in steady state conditions for some inlet pressure, steam and noncondensable gases (air) inlet mass fractions. The test section is three meters long and consists of two concentric tubes containing pressure, temperature and flow rate sensors. The internal tube, called condenser, contains steam-air mixture flow and external tube is a counter current cooler with water flow rate at low temperature. This test section was modeled and simulations were performed with RELAP5 code. Experimental tests were carried out for 200 to 400 kPa inlet pressure and 5, 10, 15 and 20% of inlet air mass fractions. Comparisons between experimental data and simulation results are presented for 200 and 400 kPa pressure conditions and showed good agreement. However, for 400 kPa inlet steam pressure and inlet air mass fractions above 5%, the simulated temperatures are lower than the experimental data at the final third from the inlet condenser tube, indicating a code overestimation of heat transfer coefficient. New correlations for heat transfer coefficient in these steam-air conditions must be theoretical and experimentally studied and implemented in RELAP5 code for better representing the condensation phenomena. (author)

  3. RELAP5/MOD3.2 Sensitivity Analysis Using OECD/NEA ROSA-2 Project 17% Cold Leg Intermediate-break LOCA Test Data

    International Nuclear Information System (INIS)

    An experiment simulating a PWR intermediate-break loss-of-coolant accident (IBLOCA) with 17% break at cold leg was conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the experiment, core dryout took place due to rapid drop in the core liquid level before loop seal clearing (LSC). Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to counter-current flow limiting (CCFL) by high velocity vapor flow, causing further decrease in the core liquid level. The post-test analysis by RELAP5/MOD3.2.1.2 code revealed that cladding surface temperature of simulated fuel rods was under-predicted due to later major core uncovery than in the experiment. Key phenomena and related important parameters, which may affect the core liquid level behavior and thus the cladding surface temperature, were selected based on the LSTF test data analysis and post-test analysis results. The post-test analysis conditions were considered as 'Base Case', for sensitivity analysis to study the causes of uncertainty in best estimate methodology. The RELAP5 sensitivity analysis was performed by changing the important parameters relevant to the key phenomena within the ranges to investigate influences of the parameters onto the cladding surface temperature. It was confirmed that both constant C of Wallis CCFL correlation at the core exit and gas-liquid inter-phase drag in the core, as parameters that need to consider for the evaluation of safety margin, are more sensitive to the cladding surface temperature than other chosen parameters. (authors)

  4. Synthesis of hierarchical porous carbon monoliths with incorporated metal-organic frameworks for enhancing volumetric based CO₂ capture capability.

    Science.gov (United States)

    Qian, Dan; Lei, Cheng; Hao, Guang-Ping; Li, Wen-Cui; Lu, An-Hui

    2012-11-01

    This work aims to optimize the structural features of hierarchical porous carbon monolith (HCM) by incorporating the advantages of metal-organic frameworks (MOFs) (Cu₃(BTC)₂) to maximize the volumetric based CO₂ capture capability (CO₂ capacity in cm³ per cm³ adsorbent), which is seriously required for the practical application of CO₂ capture. The monolithic HCM was used as a matrix, in which Cu₃(BTC)₂ was in situ synthesized, to form HCM-Cu₃(BTC)₂ composites by a step-by-step impregnation and crystallization method. The resulted HCM-Cu₃(BTC)₂ composites, which retain the monolithic shape and exhibit unique hybrid structure features of both HCM and Cu₃(BTC)₂, show high CO₂ uptake of 22.7 cm³ cm⁻³ on a volumetric basis. This value is nearly as twice as the uptake of original HCM. The dynamic gas separation measurement of HCM-Cu₃(BTC)₂, using 16% (v/v) CO₂ in N₂ as feedstock, illustrates that CO₂ can be easily separated from N₂ under the ambient conditions and achieves a high separation factor for CO₂ over N₂, ranging from 67 to 100, reflecting a strongly competitive CO₂ adsorption by the composite. A facile CO₂ release can be realized by purging an argon flow through the fixed-bed adsorber at 25 °C, indicating the good regeneration ability.

  5. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  6. Effect on code predictions by changing the code version of Relap5 on SBLOCA for Test 9.1B in BETHSY test facility

    Energy Technology Data Exchange (ETDEWEB)

    Dubey, S.K.; Gupta, S.K. [SADD, Atomic Energy Regulatory Board, Anushaktinagar, Mumbai (India); Petruzzi, A.; Giannotti, W.; D' Auria, F. [Pisa Univ., DIMNP (Italy)

    2007-07-01

    This paper deals with the use of the new version of Relap-5 code on Sbloca (small break loss of coolant accident) for test 9.1 in the Bethsy integral test facility. In this analysis an updated version of the best estimate code Relap5/mod.3.3 has been used. In this code options are available for critical flow at the junction, modified Henry Fauske model (with only one discharge coefficient and thermal non equilibrium constant), and original model (with sub-cooled, two-phase and superheated discharge coefficient). Post test analyses have been carried out. This analysis has been carried out with all the procedure lead by Uncertainty Methodology based on Accuracy Extrapolation (UMAE). In order to achieve a qualified test facility nodalization both 'steady state level' and 'on transient level' qualifications are demonstrated. It is concluded that overall qualitative and quantitative accuracy of code prediction (mod.3.3) are acceptable as per UMAE. However noticeable effect of change of version of code has been observed from mod.2 to mod.3.3. Discharge coefficient for the modified Henry Fauske model is very sensitive for the break modeling point of view. The thermal non equilibrium constant for break discharge modeling is not having much effect on the analysis results. Especially when the core is refilled, code under-predicts the break flow and integral break flow consequently code over-predicts primary mass inventory. It is found that with the same input deck all the significant events and phenomena for the mod.3.3 occurring about 600 s earlier with compare to mod.2 calculation. Up to refilling of core, time sequence of the entire significant events is after the experiments in mod.2 calculation whereas in mod.3.3 it is before the experimental value. With this new version of code a better prediction for clad temperature during dry out has been observed. In both version of code prediction of results are poor for the last 2500 s of transients this may

  7. Effect on code predictions by changing the code version of Relap5 on SBLOCA for Test 9.1B in BETHSY test facility

    International Nuclear Information System (INIS)

    This paper deals with the use of the new version of Relap-5 code on Sbloca (small break loss of coolant accident) for test 9.1 in the Bethsy integral test facility. In this analysis an updated version of the best estimate code Relap5/mod.3.3 has been used. In this code options are available for critical flow at the junction, modified Henry Fauske model (with only one discharge coefficient and thermal non equilibrium constant), and original model (with sub-cooled, two-phase and superheated discharge coefficient). Post test analyses have been carried out. This analysis has been carried out with all the procedure lead by Uncertainty Methodology based on Accuracy Extrapolation (UMAE). In order to achieve a qualified test facility nodalization both 'steady state level' and 'on transient level' qualifications are demonstrated. It is concluded that overall qualitative and quantitative accuracy of code prediction (mod.3.3) are acceptable as per UMAE. However noticeable effect of change of version of code has been observed from mod.2 to mod.3.3. Discharge coefficient for the modified Henry Fauske model is very sensitive for the break modeling point of view. The thermal non equilibrium constant for break discharge modeling is not having much effect on the analysis results. Especially when the core is refilled, code under-predicts the break flow and integral break flow consequently code over-predicts primary mass inventory. It is found that with the same input deck all the significant events and phenomena for the mod.3.3 occurring about 600 s earlier with compare to mod.2 calculation. Up to refilling of core, time sequence of the entire significant events is after the experiments in mod.2 calculation whereas in mod.3.3 it is before the experimental value. With this new version of code a better prediction for clad temperature during dry out has been observed. In both version of code prediction of results are poor for the last 2500 s of transients this may be due to large

  8. Validation of Atucha-2 PHWR helios and Relap5-3D model by Monte Carlo cell and core calculations - 335

    International Nuclear Information System (INIS)

    Within the framework of the Second Agreement 'Nucleoelectrica Argentina-SA - University of Pisa', a complex three dimensional (3D) neutron kinetics (NK) coupled thermal-hydraulic (TH) RELAP5-3D model of the Atucha 2 PHWR has been developed and validated. Homogenized cross section database was produced by the lattice physics code HELIOS. In order to increase the level of confidence on the results of such sophisticated models, an independent Monte Carlo code model, based on the MONTEBURNS package (MCNP5 + ORIGEN), has been set up. The scope of this activity is to obtain a systematic check of the deterministic codes results. This necessity is particularly felt in the case of Atucha-2 reactor modeling, since its own peculiarities (e.g., oblique Control Rods, Positive Void Coefficient) and since, if approved by the Argentinean Safety Authority, the RELAP53D 3D NK TH model will constitute the first application of a neutronic thermal-hydraulics coupled code techniques to a reactor licensing project. (authors)

  9. International agreement report: Assessment study of RELAP-5 MOD-2 Cycle 36.01 based on the DOEL-2 Steam Generator Tube Rupture incident of June 1979

    International Nuclear Information System (INIS)

    This report presents a code assessment study based on a real plant transient that occurred at the DOEL 2 power plant in Belgium on June 25th 1979. DOEL 2 is a two-loop WESTINGHOUSE PWR plant of 392 MWe. A steam generator tube rupture occurred at the end of a heat-up phase which initiated a plant transient which required substantial operator involvement and presented many plant phenomena which are of interest for code assessment. While real plant transients are of special importance for code validation because of the elimination of code scaling uncertainties, they introduce however some uncertainties related to the specifications of the exact initial and boundary conditions which must be reconstructed from available on-line plant recordings and on-line computer diagnostics. Best estimate data have been reconstructed for an assessment study by means of the code RELAP5/MOD2/CYCLE 36.01. Because of inherent uncertainties in the plant data, the assessment work is focussed on phenomena whereby the comparison between plant data and computer data is based more on trends than on absolute values. Such approach is able to uncover basic code weaknesses and strengths which can contribute to a better understanding of the code potential

  10. Assessment of RELAP5/MOD2 against a pressurizer spray valve inadverted fully opening transient and recovery by natural circulation in Jose Cabrera Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Arroyo, R.; Rebollo, L. [Union Electrica, SA, Madrid (Spain)

    1993-06-01

    This document presents the comparison between the simulation results and the plant measurements of a real event that took place in JOSE CABRERA nuclear power plant in August 30th, 1984. The event was originated by the total, continuous and inadverted opening of the pressurizer spray valve PCV-400A. JOSE CABRERA power plant is a single loop Westinghouse PWR belonging to UNION ELECTRICA FENOSA, S.A. (UNION FENOSA), an Spanish utility which participates in the International Code Assessment and Applications Program (ICAP) as a member of UNIDAD ELECTRICA, S.A. (UNESA). This is the second of its two contributions to the Program: the first one was an application case and this is an assessment one. The simulation has been performed using the RELAP5/MOD2 cycle 36.04 code, running on a CDC CYBER 180/830 computer under NOS 2.5 operating system. The main phenomena have been calculated correctly and some conclusions about the 3D characteristics of the condensation due to the spray and its simulation with a 1D tool have been got.

  11. Reactor inlet header critical break identification and analysis for KAPP-3 and 4 using computer code RELAP-5/Mod.3.2

    International Nuclear Information System (INIS)

    Kakrapar Atomic Power Project units-3 and 4 (KAPP-3 and 4) are 700 MWe Pressurized Heavy Water Reactors (PHWR) are presently under construction. This paper presents the identification of critical break in Reactor Inlet Header and its analysis performed, for KAPP-3 and 4 as a part of safety studies to investigate the plant behavior. The limiting/critical break size at Reactor Inlet Header is identified by considering the peak sheath temperature during the Loss of coolant accident. System thermal hydraulics code RELAP-5/MOD3.2 has been used for the analysis. Here the overall thermal hydraulics of the plant along with various control systems, trip and actuation logics have been simulated. High pressure accumulators and low pressure recirculation system of emergency core cooling system are modeled. The modeling of secondary system includes modeling of Atmospheric Steam Discharge Valves (ASDVs), Safety Relief Valves (SRVs), Condensate Steam Discharge Valves (CSDVs), and Governor Valves, the U-tubes of the steam generator, the riser, the separator and the steam drum. Using this model, critical break size in the Reactor Outlet Header was identified and consequence of the event on maximum peak clad temperature and core parameters were evaluated. Following postulated accidents, the event progression and the variations of different parameters like different Header pressures, mass flow rate in the core, fuel clad temperature and rate of discharge from break etc have been studied. (author)

  12. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    Science.gov (United States)

    McCann, Larry D.

    2007-01-01

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  13. Transient characteristic analyses of ex-vessel coolant pipe break for Chinese helium-cooled solid breeder TBM based on RELAP5 code

    International Nuclear Information System (INIS)

    Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system and secondary cooling water system was modeled and thermal-hydraulic behavior and safety performance of the system were assessed using the RELAP5/MOD3.4 code. According to the accident sequences of ITER accident analysis specification for TBM, the transient analysis of the design basis ex-vessel coolant pipe break accident was carried out. The influences of different break locations, leak areas and plasma shutdown processes on the first wall of TBM were compared. The results indicate that it is much more danger when the pipe break occurs at the downstream side of the helium circulator compared with that at upstream side. The results also show that the accident consequence is worse in case of smaller area break than that in case of larger area break. In case of much more severe accident that the ex-vessel break leads to the break of TBM the first wall, the results reveal that the decay heat can be removed to cool down TBM by natural circulation and radiation. The first wall melting can be avoided if the method to shutdown plasma within 3 seconds in case of ex-vessel break is adopted. (authors)

  14. Assessment of RELAP5/MOD2 against a pressurizer spray valve inadverted fully opening transient and recovery by natural circulation in Jose Cabrera Nuclear Station

    International Nuclear Information System (INIS)

    This document presents the comparison between the simulation results and the plant measurements of a real event that took place in JOSE CABRERA nuclear power plant in August 30th, 1984. The event was originated by the total, continuous and inadverted opening of the pressurizer spray valve PCV-400A. JOSE CABRERA power plant is a single loop Westinghouse PWR belonging to UNION ELECTRICA FENOSA, S.A. (UNION FENOSA), an Spanish utility which participates in the International Code Assessment and Applications Program (ICAP) as a member of UNIDAD ELECTRICA, S.A. (UNESA). This is the second of its two contributions to the Program: the first one was an application case and this is an assessment one. The simulation has been performed using the RELAP5/MOD2 cycle 36.04 code, running on a CDC CYBER 180/830 computer under NOS 2.5 operating system. The main phenomena have been calculated correctly and some conclusions about the 3D characteristics of the condensation due to the spray and its simulation with a 1D tool have been got

  15. RELAP5/MOD2 analysis of a postulated ''cold leg SBLOCA'' simultaneous to a ''total black-out'' event in the Jose Cabrera Nuclear Station

    International Nuclear Information System (INIS)

    Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ''lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic ampersand Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumption of a ''total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ''time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested

  16. RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. [Union Electrica, SA, Madrid (Spain)

    1992-04-01

    Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.

  17. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2014-01-01

    Full Text Available RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW injection into the secondary-side of both steam generators (SGs as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses.

  18. Analysis of the UPTF Separate Effects Test 11 (steam-water counter-current flow in the broken loop hot leg) using RELAP5/MOD2

    International Nuclear Information System (INIS)

    RELAP5/MOD2 predictions of countercurrent flow limitation in the UPTF hot leg separate effects Test (test 11) are compared with the experimental data. The code underestimates, by a factor of more than three, the gas flow necessary to prevent liquid runback from the steam generator, and this is shown to be due to an oversimplified flow-regime map which does not allow the possibility of stratified flow in the hot leg riser. The predicted countercurrent flow is also shown to depend, wrongly, on the depth of liquid in the steam generator plenum. The same test is also modelled using a version of the code in which stratified flow in the riser is made possible. The gas flow needed to prevent liquid runback is then predicted quite well, but at all lower gas flows the code predicts that the flow is completely unrestricted - i.e. liquid flows between full flow and zero flow are not predicted. This is shown to happen because the code cannot calculate correctly the liquid level in the hot leg, mainly because of a numerical effect of upwind donoring in the momentum flux terms of the code's basic equations. It is also shown that the code cannot model the considerable effect of the ECCS injection pipe (which runs inside the hot leg) on the liquid level. (author)

  19. RELAP5/MOD3.2 re-analysis and accuracy quantification of LOFT experiment L2-5

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A.; Giannotti, W.; D' Auria, F. [Universita di Pisa, Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa (Italy)]. E-mail: axp46@psu.edu; w.giannotti@ing.unipi.it; dauria@ing.unipi.it

    2004-07-01

    The paper presents the activity performed at University of Pisa in the framework of the participation to the Phase II of the BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme. This activity has been promoted by the Working Group on Accident Management and Analysis (GAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI). The scope of the Programme is to perform Large Break Loss-Of-Coolant Accident (LBLOCA) analyses making reference to experimental data and to a Nuclear Power Plant (NPP) in order to address the issue of 'the capabilities of computational tools' including scaling/uncertainty analysis. The justification for such an activity comes from the consideration that a wide spectrum of uncertainty methods applied to Best Estimate codes exist and are used in research laboratories, but their practicability and/or validity is not sufficiently established to support general use of the codes and acceptance by industry and safety authorities. The consideration of the Best Estimate codes and uncertainty evaluation for Design Basis Accident (DBA), by itself, shows the safety significance of the proposed activity. (author)

  20. RELAP5/MOD2.5 analysis of the HFBR [High Flux Beam Reactor] for a loss of power and coolant accident

    International Nuclear Information System (INIS)

    A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs

  1. Transient Characteristic Analyses of Ex-vessel Coolant Pipe Break for Chinese Helium-cooled Solid Breeder TBM Based on RELAP5 Code%基于RELAP5的中国氦冷固态包层真空室外破口瞬态特性分析

    Institute of Scientific and Technical Information of China (English)

    王杰; 苏光辉; 田文喜; 秋穗正; 向斌; 张国书; 冯开明

    2013-01-01

    利用RELAP5/MOD3.4对中国氦冷固态包层、氦气冷却剂回路和二次侧水冷系统进行建模和系统热工水力安全评价.依据ITER事故分析制定的事故序列,对设计基准真空室外破口进行了瞬态分析,并对比了不同破口位置、面积和停堆方式对第一壁的影响.结果表明:真空室外破口发生在风机的下游较上游危险,且小破口较大破口更危险;若真空室外破口同时包层第一壁破口,也可通过自然循环和辐射换热带走衰变热冷却包层;真空室外破口事故中采用聚变停堆系统的3 s停堆方式,可避免第一壁熔化.%Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system and secondary cooling water system was modeled and thermal-hydraulic behavior and safety performance of the system were assessed using the RELAP5/MOD3.4 code.According to the accident sequences of ITER accident analysis specification for TBM,the transient analysis of the design basis ex-vessel coolant pipe break accident was carried out.The influences of different break locations,leak areas and plasma shutdown processess on the first wall of TBM were compared.The results indicate that it is much more danger when the pipe break occurs at the downstream side of the helium circulator compared with that at upstream side.The results also show that the accident consequence is worse in case of smaller area break than that in case of larger area break.In case of much more severe accident that the ex-vessel break leads to the break of TBM the first wall,the results reveal that the decay heat can be removed to cool down TBM by natural circulation and radiation.The first wall melting can be avoided if the method to shutdown plasma within 3 seconds in case of ex-vessel break is adopted.

  2. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor; Analisis para el acoplamiento del codigo NESTLE para la cinetica tridimensional del nucleo al codigo avanzado de sistemas termo-hidraulicos, RELAP5/SCDAPSIM y su aplicacion al reactor de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J.H.; Nunez C, A. [CNSNS, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico); Chavez M, C. [UNAM, Facultad de Ingenieria, DEPFI Campus Morelos (Mexico)]. E-mail: hsalazar22@prodigy.net.mx

    2004-07-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  3. Estudio termodinámico y de degradación en un transitorio de Blackout con el código RELAP5/SCDAP de una central genérica de agua en ebullición

    OpenAIRE

    Monset Cabré, Lluís

    2011-01-01

    Esta memoria incluye una primera parte descriptiva del código de cálculo empleado (RELAP5/SCDAP) para la realización de las simulaciones, del tipo de planta nuclear en la que éstas se realizarán (BWR) y del modelo utilizado. Posteriormente se introducirá el escenario estudiado. Es un caso de accidente severo en el que, partiendo del funcionamiento en régimen estacionario de la planta, se iniciará un transitorio de Total Station Blackout. És decir, una pérdida de subministro elé...

  4. Capability of coupled 3-D neutronics/thermalhydraulic models to simulate spatial-time effects

    International Nuclear Information System (INIS)

    Last advancements in computer technology made possible the incorporation on full three-dimensional reactor core model into system transient codes. Best-estimate simulations of interactions between reactor core behavior and plant dynamics have been allowed with 3D neutronics/thermalhydraulic coupled codes. Among these codes, the RELAP5-3D has been applied to the Main Steam Line Break accident to perform three-dimensional core behavior analysis. The advantage of using a 3-D neutronics/thermalhydraulic codes is more evident in the study of strongly asymmetric transient for which simple neutron point kinetic and 1-D thermalhydraulic models are not able to provide an acceptable physical representation of the phenomena that occur in the core. The main objective of this document is to demonstrate the capability to simulated complex spatial-time effects with 3-D coupled codes. Different core nodalizations and coupling schemes have been set up. This has shown that the methodology adopted and the computational tools allow accounting for different detail levels in the core representation. (author)

  5. Incorporation of VSV-G produces fusogenic plasma membrane vesicles capable of efficient transfer of bioactive macromolecules and mitochondria.

    Science.gov (United States)

    Lin, Hao-Peng; Zheng, De-Jin; Li, Yun-Pan; Wang, Na; Chen, Shao-Jun; Fu, Yu-Cai; Xu, Wen-Can; Wei, Chi-Ju

    2016-06-01

    The objective of this study was to determine if plasma membrane vesicles (PMVs) could be exploited for efficient transfer of macro-biomolecules and mitochondria. PMVs were derived from mechanical extrusion, and made fusogenic (fPMVs) by incorporating the glycoprotein G of vesicular stomatitis virus (VSV-G). Confocal microscopy examination revealed that cytoplasmic proteins and mitochondria were enclosed in PMVs as evidenced by tracing with cytoplasmically localized and mitochondria-targeted EGFP, respectively. However, no fluorescence signal was detected in PMVs from cells whose nucleus was labeled with an EGFP-tagged histone H2B. Consistently, qRT-PCR measurement showed that mRNA, miRNA and mitochondrial DNA decreased slightly; while nuclear DNA was not measureable. Further, Western blot analysis revealed that cytoplasmic and membrane-bound proteins fell inconspicuously while nuclear proteins were barely detecsle. In addition, fPMVs carrying cytoplamic DsRed proteins transduced about ~40 % of recipient cells. The transfer of protein was further confirmed by using the inducible Cre/loxP system. Mitochondria transfer was found in about 20 % recipient cells after incubation with fPMVs for 5 h. To verify the functionalities of transferred mitochondria, mitochodria-deficient HeLa cells (Rho0) were generated and cultivated with fPMVs. Cell enumeration demonstrated that adding fPMVs into culture media stimulated Rho0 cell growth by 100 % as compared to the control. Lastly, MitoTracker and JC-1 staining showed that transferred mitochondria maintained normal shape and membrane potential in Rho0 cells. This study established a time-saving and efficient approach to delivering proteins and mitochondria by using fPMVs, which would be helpful for finding a cure to mitochondria-associated diseases. Graphical abstract Schematic of the delivery of macro-biomolecules and organelles by fPMVs. VSV-G-expressing cells were extruded through a 3 μm polycarbonate membrane filter to

  6. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm2, simulated with RELAP5 code

    International Nuclear Information System (INIS)

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm2-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  7. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  8. Estudio termodinámico y estructural de un escenario transitorio de Blackout en la planta nuclear de Ascó II con el código RELAP5/SCDAP

    OpenAIRE

    Weill , Vincent

    2005-01-01

    Esta memoria incluye una primera parte descriptiva del modelo de planta de Ascó II, modelo híbrido realizado utilizando el código de cálculo RELAP5/SCDAP. El término híbrido significa que se trata de dos aspectos del diseño de la planta nuclear: la descripción de la vasija con su nodalización específica SCDAP y la descripción de la parte hidrodinámica, es decir de todos los otros componentes, con una nodalización RELAP. Una vez el modelo descrito, se introducirá el escenario es...

  9. ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2016-01-01

    Full Text Available Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility. The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

  10. RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Electrica, SA, Madrid (Spain))

    1992-04-01

    Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumption of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.

  11. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor

    International Nuclear Information System (INIS)

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  12. Analysis of experiments performed at University of Hannover with Relap5/Mod2 and Cathare codes on fluid dynamic effects in the fuel element top nozzle area during refilling and reflooding

    International Nuclear Information System (INIS)

    The experimental data of flooding and CCFL in the fuel element top nozzle area collected at the University of Hannover have been analyzed with RELAP5/MOD2 and CATHARE V.1.3 codes. Preliminary sensitivity calculations have been performed to evaluate the influence of various parameters and code options on the results. However, an a priori rational assessment procedure has been performed for those parameters non specific in experimental data (e.g. energy loss coefficients in flow restrictions). This procedure is based on single phase flow pressure drops and no further tuning has been performed to fit experimental data. The reported experimental data and some others demonstrate the complex relation-ship among the involved physical quantities (film thickness, pressure drop etc.) even in a simple geometrical condition with well defined boundary conditions. In the application of the two advanced codes to the selected CCFL experiments it appears that sophisticated models do not simulate satisfactorily the measured phenomena mainly when situations similar to nuclear reactors are dealt with (rod bundles). This result should be evaluated considering that: - dimensional phenomena occurring in flooding experiments are not well reproducible with one dimensional models implemented in the two codes; - a rational and reproducible procedure has been used to fix some boundary conditions (K-tuning); there is the evidence that more tuning can be used to get results closer to the experimental ones in each specific situation; - the uncertainty bands in measured experimental results are not (entirely) specified. The work performed demonstrated that further applications to CCFL experiments of present codes appear to be unuseful. New models should be tested and implemented before any attempt to reproduce CCFL in experimental facilities by system codes

  13. CRM Systems with Social Networking Capabilities: The Value of Incorporating a CRM 2.0 System in Sales/Marketing Education

    Science.gov (United States)

    Wang, Xin; Dugan, Riley; Sojka, Jane

    2013-01-01

    Implementation of a customer relationship management (CRM) 2.0 system can provide both a valuable pedagogical tool and a needed skill set in a marketing and sales curriculum. A CRM 2.0 system incorporated in the sales and marketing curriculum can help manage relationships between students, practitioners, and faculty while teaching students a…

  14. Models for describing the behaviour of light water reactors in serious accidents for the programs SCDAP/RELAP5, ATHLET/SA, CATHARE/ICARE, MELCOR etc.. First technical report on BMFT-sponsored research project 1500 831 7: Comparative assessment of different computer codes for severe accident analysis, contribution to the ATHLET/CD code development

    International Nuclear Information System (INIS)

    Within the scope of the project BMFT No. 15008317 entitled ''Comparative Assessment of Different Computer Codws for Severe Accident Analysis, Contribution to the ATHLET/SA-Code Development'' the codes ATHLET/SA, CATHARE/ICARE, MELCOR and SCDAP/RELAP5 are investigated. Emphasis is put on a comparison and an assessment of the governing modelling features implemented and operating in the codes under consideration. The codes are evaluated and compared on the base of selected experiments (especially the CORA experimental program of the Karlsruhe Research Center) and relevant severe accident scenarios. The present report is a reference study dealing with the governing models implemented in the severe accident codes SCDAP/RELAP5, ATHLET/SA, CATHARE/ICARE, MELCOR, KESS-III, MAAP and MELPROG/TRAC. Emphaisis is laid on the following models (molstly implemented in form of modules in the respective codes) dealing with: - thermal hydraulics; - heat generation and heat structures; - Radiation heat transfer; - mechanical (rod) behaviour; - core heatup, meltdown and relocation; - chemical reaction; - fission product release and transport; - material properties; - specific components. (orig.)

  15. Dynamic Capabilities and Performance

    DEFF Research Database (Denmark)

    Wilden, Ralf; Gudergan, Siegfried P.; Nielsen, Bo Bernhard;

    2013-01-01

    Dynamic capabilities are widely considered to incorporate those processes that enable organizations to sustain superior performance over time. In this paper, we argue theoretically and demonstrate empirically that these effects are contingent on organizational structure and the competitive...... are contingent on the competitive intensity faced by firms. Our findings demonstrate the performance effects of internal alignment between organizational structure and dynamic capabilities, as well as the external fit of dynamic capabilities with competitive intensity. We outline the advantages of PLS...

  16. Capability Paternalism

    NARCIS (Netherlands)

    Claassen, R.J.G.

    2014-01-01

    A capability approach prescribes paternalist government actions to the extent that it requires the promotion of specific functionings, instead of the corresponding capabilities. Capability theorists have argued that their theories do not have much of these paternalist implications, since promoting c

  17. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 4. Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5

    International Nuclear Information System (INIS)

    The purpose of this work is to discuss the methods developed in our three-dimensional (3-D) pressurized water reactor (PWR) SIMTRAN Core Dynamics code and its coupling to the RELAP-5 system code for general transient and safety analysis, as well as its demonstration application to the Nuclear Energy Agency/Organization for Economic Cooperation and Development (NEA/OECD) Benchmark on Main Steam Line Break (MSLB), cosponsored by the U.S. Nuclear Regulatory Commission (NRC) and other regulatory institutions. In particular, our work has been supported by the Spanish Consejo de Seguridad Nuclear (CSN) under a CSN research project. SIMTRAN is our 3-D PWR core dynamics code,1 which has been under development and validation for ∼10 yr (Refs. 1, 2, and 3). It was developed as a single code merge, with data sharing through standard FORTRAN commons, of our SIMULA 3-D neutronics nodal code and the COBRA-IIIC/MIT-2 multichannel, with cross-flows, thermal-hydraulics (T-H) code. Both codes solve the 3-D neutronic and T-H fields with maximum implicitness, using direct and iterative methods for the inversion of the linearized systems. SIMULA uses synthetic coarse-mesh discontinuity factors, in the XY directions, pre-calculated by two-dimensional (2-D) pin-by-pin two-group diffusion calculations of whole core planes, and embedded iterative one-dimensional (1-D) fine-mesh two-group diffusion solutions in the axial direction. COBRA uses direct inversion at each plane of the axial flow equations, with cross-flows updated over an outer iteration loop, for the homogenous model single-phase coolant, and finite element direct solution of the fuel rod radial temperatures. The 3-D core N-T-H coupling is done internally by a semi-implicit scheme, using a staggered alternate time mesh, where the T-H solution is done at the half of the neutronic time step (thus conserving energy by taking the power centered in the time step) and extrapolating the 3-D T-H variables over a half of the time step

  18. Capability ethics

    NARCIS (Netherlands)

    I.A.M. Robeyns (Ingrid)

    2012-01-01

    textabstractThe capability approach is one of the most recent additions to the landscape of normative theories in ethics and political philosophy. Yet in its present stage of development, the capability approach is not a full-blown normative theory, in contrast to utilitarianism, deontological theor

  19. Dynamic Capabilities

    DEFF Research Database (Denmark)

    Grünbaum, Niels Nolsøe; Stenger, Marianne

    2013-01-01

    The findings reveal a positive relationship between dynamic capabilities and innovation performance in the case enterprises, as we would expect. It was, however, not possible to establish a positive relationship between innovation performance and profitability. Nor was there any positive...... relationship between dynamic capabilities and profitability....

  20. Dynamic Capabilities

    DEFF Research Database (Denmark)

    Grünbaum, Niels Nolsøe; Stenger, Marianne

    2013-01-01

    The findings reveal a positive relationship between dynamic capabilities and innovation performance in the case enterprises, as we would expect. It was, however, not possible to establish a positive relationship between innovation performance and profitability. Nor was there any positive relation......The findings reveal a positive relationship between dynamic capabilities and innovation performance in the case enterprises, as we would expect. It was, however, not possible to establish a positive relationship between innovation performance and profitability. Nor was there any positive...... relationship between dynamic capabilities and profitability....

  1. Gossiping Capabilities

    DEFF Research Database (Denmark)

    Mogensen, Martin; Frey, Davide; Guerraoui, Rachid;

    declare a high capability in order to augment their perceived quality without contributing accordingly. We evaluate HEAP in the context of a video streaming application on a 236 PlanetLab nodes testbed. Our results shows that HEAP improves the quality of the streaming by 25% over a standard gossip......Gossip-based protocols are now acknowledged as a sound basis to implement collaborative high-bandwidth content dissemination: content location is disseminated through gossip, the actual contents being subsequently pulled. In this paper, we present HEAP, HEterogeneity Aware gossip Protocol, where...... nodes dynamically adjust their contribution to gossip dissemination according to their capabilities. Using a continuous, itself gossip-based, approximation of relative capabilities, HEAP dynamically leverages the most capable nodes by (a) increasing their fanouts (while decreasing by the same proportion...

  2. RELAP5 analysis of an EOP based on mobile pumps, at a generic VVER-1000 NPP in case of a total loss of the primary heat sink (for DBDA conditions)

    Energy Technology Data Exchange (ETDEWEB)

    Nikolaus Muellner [Institute of Risk Research (Vienna University), Tuerkenschanzstr. 17/8, A-1180 Vienna (Austria); Walter Giannotti; Francesco D' Auria [University of Pisa, Via Diotisalvi 2, 56100 Pisa (Italy)

    2005-07-01

    Full text of publication follows: The loss of the primary heat sink is one of the most relevant DBDA scenarios. If the capability to remove heat from the primary side cannot be restored, the primary side will be subjected to high pressure for a long period of time. The actuation of the ECCS will be inhibited, PS inventory will be lost, finally the core will be in dryout conditions. The EOP which is investigated in this paper proposes to utilize mobile pumps (e.g. using fire brigade trucks) to make additional sources of feedwater available and thereby restore the primary heat sink at least for a limited period of time. After a certain time (which should be determined by calculations) of this measure, the PORV-valve and/or the gas-removal system should be opened to lower the pressure in the primary side and actuate the ECCS (primary side bleed and feed mode). The calculations indicate that this procedure is capable of saving the plant, or at least extending the grace period. Three cases are presented: the effect of only primary side feed and bleed, the beneficial effect of utilising mobile pumps first, and a scenario which assumes no operator action. (authors)

  3. ENTREPRENEURIAL CAPABILITIES

    DEFF Research Database (Denmark)

    Rasmussen, Lauge Baungaard; Nielsen, Thorkild

    2003-01-01

    The aim of this article is to analyse entrepreneurship from an action research perspective. What is entrepreneurship about? Which are the fundamental capabilities and processes of entrepreneurship? To answer these questions the article includes a case study of a Danish entrepreneur and his networks...

  4. Capability approach

    DEFF Research Database (Denmark)

    Jensen, Niels Rosendal; Kjeldsen, Christian Christrup

    Lærebogen er den første samlede danske præsentation af den af Amartya Sen og Martha Nussbaum udviklede Capability Approach. Bogen indeholder en præsentation og diskussion af Sen og Nussbaums teoretiske platform. I bogen indgår eksempler fra såvel uddannelse/uddannelsespolitik, pædagogik og omsorg....

  5. RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [KOPEC, Yongin (Korea, Republic of); Kim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [KEPCO, Seoul (Korea, Republic of)

    1998-05-01

    The Reactor Coolant Gas Vent System (RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Coodown (NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown pheonomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifice size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability test, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation.

  6. RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Lim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [Korea Electric Power Cooperation, Taejon (Korea, Republic of)

    1998-12-31

    The Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown phenomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifics size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation. 6 refs., 5 figs., 1 tab. (Author)

  7. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs

  8. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K. (ed.) (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-02-01

    This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs.

  9. Layered Composite Analysis Capability

    Science.gov (United States)

    Narayanaswami, R.; Cole, J. G.

    1985-01-01

    Laminated composite material construction is gaining popularity within industry as an attractive alternative to metallic designs where high strength at reduced weights is of prime consideration. This has necessitated the development of an effective analysis capability for the static, dynamic and buckling analyses of structural components constructed of layered composites. Theoretical and user aspects of layered composite analysis and its incorporation into CSA/NASTRAN are discussed. The availability of stress and strain based failure criteria is described which aids the user in reviewing the voluminous output normally produced in such analyses. Simple strategies to obtain minimum weight designs of composite structures are discussed. Several example problems are presented to demonstrate the accuracy and user convenient features of the capability.

  10. Physicochemical characterization of surfactant incorporating vesicles that incorporate colloidal magnetite.

    Science.gov (United States)

    de Melo Barbosa, Raquel; Luna Finkler, Christine L; Bentley, Maria Vitória L B; Santana, Maria Helena A

    2013-03-01

    Drug administration through the transdermal route has optimized for the comfort of patients and easy application. However, the main limitation of transdermal drug delivery is the impermeability of the human skin. Recent advances on improvement of drug transport through the skin include elastic liposomes as a penetration enhancer. Entrapment of ferrofluids in the core of liposomes produces magnetoliposomes, which can be driven by a high-gradient magnetic field. The association of both strategies could enhance the penetration of elastic liposomes. This work relies on the preparation and characterization of elastic-magnetic liposomes designed to permeate through the skin. The incorporation of colloidal magnetite and the elastic component, octaethylene glycol laurate (PEG-8-L), in the structure of liposomes were evaluated. The capability of the elastic magnetoliposomes for permeation through nanopores of two stacked polycarbonate membranes was compared to conventional and elastic liposomes. Magnetite incorporation was dependent on vesicle diameter and size distribution as well as PEG-8-L incorporation into liposomes, demonstrating the capability of the fluid bilayer to accommodate the surfactant without disruption. On the contrary, PEG-8-L incorporation into magnetoliposomes promoted a decrease of average diameter and a lower PEG-8-L incorporation percentage as a result of reduction on the fluidity of the bilayer imparted by iron incorporation into the lipid structure. Elastic liposomes demonstrated an enhancement of the deformation capability, as compared with conventional liposomes. Conventional and elastic magnetoliposomes presented a reduced capability for deformation and permeation. PMID:23363304

  11. Building Service Provider Capabilities

    DEFF Research Database (Denmark)

    Brandl, Kristin; Jaura, Manya; Ørberg Jensen, Peter D.

    process. We find that clients influence the development of human capital capabilities and management capabilities in reciprocally produced services. While in sequential produced services clients influence the development of organizational capital capabilities and management capital capabilities....

  12. IAC - INTEGRATED ANALYSIS CAPABILITY

    Science.gov (United States)

    Frisch, H. P.

    1994-01-01

    Integration via Mesh Interpolation Coefficients), which transforms field values from one model to another; LINK, which simplifies incorporation of user specific modules into IAC modules; and DATAPAC, the National Bureau of Standards statistical analysis package. The IAC database contains structured files which provide a common basis for communication between modules and the executive system, and can contain unstructured files such as NASTRAN checkpoint files, DISCOS plot files, object code, etc. The user can define groups of data and relations between them. A full data manipulation and query system operates with the database. The current interface modules comprise five groups: 1) Structural analysis - IAC contains a NASTRAN interface for standalone analysis or certain structural/control/thermal combinations. IAC provides enhanced structural capabilities for normal modes and static deformation analysis via special DMAP sequences. IAC 2.5 contains several specialized interfaces from NASTRAN in support of multidisciplinary analysis. 2) Thermal analysis - IAC supports finite element and finite difference techniques for steady state or transient analysis. There are interfaces for the NASTRAN thermal analyzer, SINDA/SINFLO, and TRASYS II. FEMNET, which converts finite element structural analysis models to finite difference thermal analysis models, is also interfaced with the IAC database. 3) System dynamics - The DISCOS simulation program which allows for either nonlinear time domain analysis or linear frequency domain analysis, is fully interfaced to the IAC database management capability. 4) Control analysis - Interfaces for the ORACLS, SAMSAN, NBOD2, and INCA programs allow a wide range of control system analyses and synthesis techniques. Level 2.5 includes EIGEN, which provides tools for large order system eigenanalysis, and BOPACE, which allows for geometric capabilities and finite element analysis with nonlinear material. Also included in IAC level 2.5 is SAMSAN 3.1, an

  13. Mobile Test Capabilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Electrical Power Mobile Test capabilities are utilized to conduct electrical power quality testing on aircraft and helicopters. This capability allows that the...

  14. Study on Mitigating Capability of AFW System for SBO Accident%辅助给水系统对缓解全厂断电事故能力研究

    Institute of Scientific and Technical Information of China (English)

    张往锁; 曹夏昕; 阎昌琪; 陈薇

    2012-01-01

    以CPR1000核电站为研究对象,采用RELAP5/MOD3.2轻水堆瞬态分析程序,对系统进行合理简化并建模,模拟系统在全厂断电事故下的瞬态响应过程,研究全厂断电事故发生后辅助给水(AFW)的投入对缓解全厂断电事故的能力.计算结果表明:断电事故发生后,主给水丧失导致一回路压力和冷却剂平均温度在断电后6 s达到峰值;辅助给水投入约200 s后,一回路因热阱丧失而引起的温度和压力升高能有效地得到缓解,为交流电源的恢复及余热排出系统的投入赢得了更多的时间.%The transient characteristic of the simplified system for CPR1000 under the station blackout (SBO) accident was simulated by use of RELAP5/MOD3. 2 code, and the capability of mitigating accident results after auxiliary feedwater (AFW) system operating was also analyzed. The calculation results show that; As a result of the loss of the main water supply, the average temperature and the pressure of the primary system reach their maximum in 6 s after SBO accident; the increasing of temperature and pressure caused by the loss of heat sink is effectively mitigated in about 200 s after the input of auxiliary feed water, and this will win more time for the recovery of AC power and the input of residual heat removal system.

  15. Capabilities for Strategic Adaptation

    DEFF Research Database (Denmark)

    Distel, Andreas Philipp

    firms’ ability to absorb and leverage new knowledge. The third paper is an empirical study which conceptualizes top managers’ resource cognition as a managerial capability underlying firms’ resource adaptation; it empirically examines the performance implications of this capability and organizational......This dissertation explores capabilities that enable firms to strategically adapt to environmental changes and preserve competitiveness over time – often referred to as dynamic capabilities. While dynamic capabilities being a popular research domain, too little is known about what these capabilities...... empirical studies through the dynamic capabilities lens and develops propositions for future research. The second paper is an empirical study on the origins of firm-level absorptive capacity; it explores how organization-level antecedents, through their impact on individual-level antecedents, influence...

  16. Developing Alliance Capabilities

    DEFF Research Database (Denmark)

    Heimeriks, Koen H.; Duysters, Geert; Vanhaverbeke, Wim

    This paper assesses the differential performance effects of learning mechanisms on the development of alliance capabilities. Prior research has suggested that different capability levels could be identified in which specific intra-firm learning mechanisms are used to enhance a firm's alliance...... capability. However, empirical testing in this field is scarce and little is known as to what extent different learning mechanisms are indeed useful in advancing a firm's alliance capability. This paper analyzes to what extent intra-firm learning mechanisms help firms develop their alliance capability....... Differential learning may explain in what way firms yield superior returns from their alliances in comparison to competitors. The empirical results show that different learning mechanisms have different performance effects at different stages of the alliance capability development process. The main lesson from...

  17. Integrated Process Capability Analysis

    Institute of Scientific and Technical Information of China (English)

    Chen; H; T; Huang; M; L; Hung; Y; H; Chen; K; S

    2002-01-01

    Process Capability Analysis (PCA) is a powerful too l to assess the ability of a process for manufacturing product that meets specific ations. The larger process capability index implies the higher process yield, a nd the larger process capability index also indicates the lower process expected loss. Chen et al. (2001) has applied indices C pu, C pl, and C pk for evaluating the process capability for a multi-process product wi th smaller-the-better, larger-the-better, and nominal-the-best spec...

  18. Capabilities for innovation

    DEFF Research Database (Denmark)

    Nielsen, Peter; Nielsen, René Nesgaard; Bamberger, Simon Grandjean;

    2012-01-01

    Technological developments combined with increasing levels of competition related to the ongoing globalization imply that firms find themselves in dynamic, changing environments that call for dynamic capabilities. This challenges the internal human and organizational resources of firms in general...... and in particular their ability to develop firm-specific innovative capabilities through employee participation and creation of innovative workplaces. In this article, we argue that national institutional conditions can play an enhancing or hampering role in this. Especially the norms and values governing relations...... between employers and employees are expected to be of vital importance. This article will follow a resource-based perspective on developing dynamic capabilities in order to test the importance of enhancing human and organizational capabilities for innovation in firms. In particular, the article will focus...

  19. Education and Innovative Capabilities

    OpenAIRE

    A. Leiponen

    1996-01-01

    This study investigates the role of capabilities, acquired through education and on the job learning, in innovation. It is argued that education enhances learning and innovation because it provides employees with communication and interaction skills, and, more importantly, with abilities to receive, understand and utilize relevant knowledge, and solve problems. These dynamic capabilities are one of the sources of innovation. A dataset of 333 Finnish manufacturing firms is used to estimat...

  20. Present status of the 'CIAU' (Code with capability of Internal Assessment of Uncertainty) and application possibilities in the licensing process

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F.; Giannotti, W. [Pisa Univ. (Italy). Dipt. di Ingegneria Meccanica, Nucleare e della Produzione. E-mail: dauria@ing.unipi.it

    2000-07-01

    The internal assessment constitutes a desirable capacity for thermalhydraulic system codes allowing the achievement of uncertainty bands associated with any code-calculation result. Currently, the CIAU (Code with - the capability of - Internal Assessment of Uncertainty) is derived from the UMAE (Uncertainty Methodology based on the Accuracy Extrapolation), through other uncertainty methodologies can be used for the same purpose. Therefore, the system code, Relap5 in the present case, is coupled with the uncertainty methodology and constitutes the CIAU. The idea at the basis of CIAU is the identification and the characterization of standard plant statuses and the association of uncertainty to each status. One hypercube and one time interval identify the plant status. Quantity and Time uncertainties are combined for each plant status. In the present paper, the derivation of the methodology is outlined and typical results of PWR (Pressurized Water reactor) plant transients are shown bounded by the CIAU calculated uncertainty bands. Results of a bifurcation study are shown and the possibility of integrating the methodology within the licensing process is discussed. (author)

  1. Structural Capability of an Organization toward Innovation Capability

    DEFF Research Database (Denmark)

    Nielsen, Susanne Balslev; Momeni, Mostafa

    2016-01-01

    competitive advantage in the organizations is the innovation capability. The innovation capability is associated with other organizational capabilities, and many organizations have focused on the need to identify innovation capabilities.This research focuses on recognition of the structural aspect...

  2. Building server capabilities

    DEFF Research Database (Denmark)

    Adeyemi, Oluseyi

    Many western companies have moved part of their operations to China in order to take advantage of cheap resources and/or to gain access to a high potential market. Depending on motive, offshore facilities usually start either as “sales-only” of products exported by headquarters or “production......-only”, exporting parts and components back to headquarter for sales in the home country. In the course of time, the role of offshore subsidiaries in a company’s operations network tends to change and, with that, the capabilities, of the subsidiaries. Focusing on Danish subsidiaries in China, the objective...... of this project is to identify and explain trajectories of offshore subsidiary capability development. The domain of inquiry is defined as value chain capabilities. Given the nature of this objective the chief methodology is longitudinal, partly retrospective, partly real-time, case studies....

  3. Implementation of Extreme STOL Capability in Cruise Efficient Aircraft Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Aerotonomy, Incorporated and the Georgia Tech Research Institute (GTRI), will develop enabling technologies for an aircraft that is capable of Extreme Short Takeoff...

  4. Capability Handbook- offline metrology

    DEFF Research Database (Denmark)

    Islam, Aminul; Marhöfer, David Maximilian; Tosello, Guido;

    This offline metrological capability handbook has been made in relation to HiMicro Task 3.3. The purpose of this document is to assess the metrological capability of the HiMicro partners and to gather the information of all available metrological instruments in the one single document. It provides...... a quick overview of what is possible today by the state of the art, what the HiMicro consortium can do and what metrological requirements we have concerning the HiMicro industrial demonstrators....

  5. Management Innovation Capabilities

    DEFF Research Database (Denmark)

    Harder, Mie

    , the paper introduces the concept of management innovation capabilities which refers to the ability of a firm to purposefully create, extend and modify its managerial resource base to address rapidly changing environments. Drawing upon behavioral theory of the firm and the dynamic capabilities framework......Management innovation is the implementation of a new management practice, process, technique or structure that significantly alters the way the work of management is performed. This paper presents a typology categorizing management innovation along two dimensions; radicalness and complexity. Then......, the paper proposes a model of the foundations of management innovation. Propositions and implications for future research are discussed....

  6. Metrology Measurement Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Glen E. Gronniger

    2007-10-02

    This document contains descriptions of Federal Manufacturing & Technologies (FM&T) Metrology capabilities, traceability flow charts, and the measurement uncertainty of each measurement capability. Metrology provides NIST traceable precision measurements or equipment calibration for a wide variety of parameters, ranges, and state-of-the-art uncertainties. Metrology laboratories conform to the requirements of the Department of Energy Development and Production Manual Chapter 13.2, ANSI/ISO/IEC ANSI/ISO/IEC 17025:2005, and ANSI/NCSL Z540-1. FM&T Metrology laboratories are accredited by NVLAP for the parameters, ranges, and uncertainties listed in the specific scope of accreditation under NVLAP Lab code 200108-0. See the Internet at http://ts.nist.gov/Standards/scopes/2001080.pdf. These parameters are summarized. The Honeywell Federal Manufacturing & Technologies (FM&T) Metrology Department has developed measurement technology and calibration capability in four major fields of measurement: (1) Mechanical; (2) Environmental, Gas, Liquid; (3) Electrical (DC, AC, RF/Microwave); and (4) Optical and Radiation. Metrology Engineering provides the expertise to develop measurement capabilities for virtually any type of measurement in the fields listed above. A strong audit function has been developed to provide a means to evaluate the calibration programs of our suppliers and internal calibration organizations. Evaluation includes measurement audits and technical surveys.

  7. Exploration Medical Capability

    Science.gov (United States)

    Watkins, Sharmila; Baumann, David; Wu, Jimmy; Barsten, Kristina

    2010-01-01

    Exploration Medical Capability (ExMC) is an element of NASA's Human Research Program (HRP). ExMC's goal is to address the risk of the Inability to Adequately Recognize or Treat an Ill or Injured Crewmember. This poster highlights the approach ExMC has taken to address this goal and our current areas of interest. The Space Medicine Exploration Medical Condition List (SMEMCL) was created to identify medical conditions of concern during exploration missions. The list was derived from space flight medical incidents, the shuttle medical checklist, the International Space Station medical checklist, and expert opinion. The conditions on the list were prioritized according to mission type by a panel comprised of flight surgeons, physician astronauts, engineers, and scientists. From the prioritized list, the ExMC element determined the capabilities needed to address the medical conditions of concern. Where such capabilities were not currently available, a gap was identified. The element s research plan outlines these gaps and the tasks identified to achieve the desired capabilities for exploration missions. This poster is being presented to inform the audience of the gaps and tasks being investigated by ExMC and to encourage discussions of shared interests and possible future collaborations.

  8. Capabilities for Intercultural Dialogue

    Science.gov (United States)

    Crosbie, Veronica

    2014-01-01

    The capabilities approach offers a valuable analytical lens for exploring the challenge and complexity of intercultural dialogue in contemporary settings. The central tenets of the approach, developed by Amartya Sen and Martha Nussbaum, involve a set of humanistic goals including the recognition that development is a process whereby people's…

  9. ISOPHOT - Capabilities and performance

    DEFF Research Database (Denmark)

    Lemke, D.; Klaas, U.; Abolins, J.;

    1996-01-01

    ISOPHOT covers the largest wavelength range on ISO from 2.5 to 240 mu m. Its scientific capabilities include multi filter and multi-aperture photometry, polarimetry, imaging and spectrophotometry. All modes can optionally include a focal plane chopper. The backbone of the photometric calibration ...

  10. Capitalizing on capabilities.

    Science.gov (United States)

    Ulrich, Dave; Smallwood, Norm

    2004-06-01

    By making the most of organizational capabilities--employees' collective skills and fields of expertise--you can dramatically improve your company's market value. Although there is no magic list of proficiencies that every organization needs in order to succeed, the authors identify 11 intangible assets that well-managed companies tend to have: talent, speed, shared mind-set and coherent brand identity, accountability, collaboration, learning, leadership, customer connectivity, strategic unity, innovation, and efficiency. Such companies typically excel in only three of these capabilities while maintaining industry parity in the other areas. Organizations that fall below the norm in any of the 11 are likely candidates for dysfunction and competitive disadvantage. So you can determine how your company fares in these categories (or others, if the generic list doesn't suit your needs), the authors explain how to conduct a "capabilities audit," describing in particular the experiences and findings of two companies that recently performed such audits. In addition to highlighting which intangible assets are most important given the organization's history and strategy, this exercise will gauge how well your company delivers on its capabilities and will guide you in developing an action plan for improvement. A capabilities audit can work for an entire organization, a business unit, or a region--indeed, for any part of a company that has a strategy to generate financial or customer-related results. It enables executives to assess overall company strengths and weaknesses, senior leaders to define strategy, midlevel managers to execute strategy, and frontline leaders to achieve tactical results. In short, it helps turn intangible assets into concrete strengths. PMID:15202293

  11. How do organizational capabilities shape industry dynamics?

    OpenAIRE

    Corsino, Marco; Gabriele, Roberto; Zaninotto, Enrico

    2009-01-01

    This paper aims to reconcile the logic behind stochastic models of firm growth and the notion of organizational capabilities as drivers of economic performance. In the proposed behavioral model of bounded rational firms, two mechanisms drive growth: independent stochastic growth of individual opportunities and the process by which firms capture new opportunities. To extend the stochastic framework, this research incorporates behavioral assumptions about the interactions between the firm and t...

  12. Atmospheric Release Advisory Capability

    International Nuclear Information System (INIS)

    The Atmospheric Release Advisory Capability (ARAC) project is a Department of Energy (DOE) sponsored real-time emergency response service available for use by both federal and state agencies in case of a potential or actual atmospheric release of nuclear material. The project, initiated in 1972, is currently evolving from the research and development phase to full operation. Plans are underway to expand the existing capability to continuous operation by 1984 and to establish a National ARAC Center (NARAC) by 1988. This report describes the ARAC system, its utilization during the past two years, and plans for its expansion during the next five to six years. An integral part of this expansion is due to a very important and crucial effort sponsored by the Defense Nuclear Agency to extend the ARAC service to approximately 45 Department of Defense (DOD) sites throughout the continental US over the next three years

  13. Building Server Capabilities

    DEFF Research Database (Denmark)

    Adeyemi, Oluseyi

    2013-01-01

    Many western companies have moved part of their operations to China in order to take advantage of cheap resources and/or to gain access to a high potential market. Depending on motive, offshore facilities usually start either as “sales-only” of products exported by headquarters or “production......-only”, exporting parts and components back to headquarter for sales in the home country. In the course of time, the role of offshore subsidiaries in a company’s operations network tends to change and, with that, the capabilities, of the subsidiaries. Focusing on Danish subsidiaries in China, the objective...... of this project is to identify and explain trajectories of offshore subsidiary capability development. Given the nature of this objective the chief methodology is longitudinal, partly retrospective, partly real-time, case studies....

  14. Management Innovation Capabilities

    OpenAIRE

    Harder, Mie

    2011-01-01

    Management innovation is the implementation of a new management practice, process, technique or structure that significantly alters the way the work of management is performed. This paper presents a typology categorizing management innovation along two dimensions; radicalness and complexity. Then, the paper introduces the concept of management innovation capabilities which refers to the ability of a firm to purposefully create, extend and modify its managerial resource base to add...

  15. Knowledge Management Capabilities Rubrics

    OpenAIRE

    Azizah B.A. Rahman; Sara Hassani

    2011-01-01

    Problem statement: Recently researchers discerned the vitality and importance of Knowledge Management Capabilities (KMC) evaluation in organizations. In fact evaluation of KMC helps to prevent failure in Knowledge Management (KM) projects. Approach: One of the most popular methods in the phase of evaluating KMC is Fuzzy method which evaluates seven attributes of KMC. Fuzzy needs KM experts to give their opinion about these attributes as input data. However in some organizations these experts ...

  16. Atmospheric release advisory capability

    International Nuclear Information System (INIS)

    The ARAC system (Atmospheric Release Advisory Capability) is described. The system is a collection of people, computers, computer models, topographic data and meteorological input data that together permits a calculation of, in a quasi-predictive sense, where effluent from an accident will migrate through the atmosphere, where it will be deposited on the ground, and what instantaneous and integrated dose an exposed individual would receive

  17. Group Capability Model

    Science.gov (United States)

    Olejarski, Michael; Appleton, Amy; Deltorchio, Stephen

    2009-01-01

    The Group Capability Model (GCM) is a software tool that allows an organization, from first line management to senior executive, to monitor and track the health (capability) of various groups in performing their contractual obligations. GCM calculates a Group Capability Index (GCI) by comparing actual head counts, certifications, and/or skills within a group. The model can also be used to simulate the effects of employee usage, training, and attrition on the GCI. A universal tool and common method was required due to the high risk of losing skills necessary to complete the Space Shuttle Program and meet the needs of the Constellation Program. During this transition from one space vehicle to another, the uncertainty among the critical skilled workforce is high and attrition has the potential to be unmanageable. GCM allows managers to establish requirements for their group in the form of head counts, certification requirements, or skills requirements. GCM then calculates a Group Capability Index (GCI), where a score of 1 indicates that the group is at the appropriate level; anything less than 1 indicates a potential for improvement. This shows the health of a group, both currently and over time. GCM accepts as input head count, certification needs, critical needs, competency needs, and competency critical needs. In addition, team members are categorized by years of experience, percentage of contribution, ex-members and their skills, availability, function, and in-work requirements. Outputs are several reports, including actual vs. required head count, actual vs. required certificates, CGI change over time (by month), and more. The program stores historical data for summary and historical reporting, which is done via an Excel spreadsheet that is color-coded to show health statistics at a glance. GCM has provided the Shuttle Ground Processing team with a quantifiable, repeatable approach to assessing and managing the skills in their organization. They now have a common

  18. Capabilities and Special Needs

    DEFF Research Database (Denmark)

    Kjeldsen, Christian Christrup

    into international consideration in relation to the implementation of the UN convention on the rights of persons with disabilities. As for the theoretical basis, the research makes use of the sociological open-ended and relational concepts of Pierre Bourdieu and the normative yardstick of the Capability Approach....... An approach that has been formulated by Amartya Sen and Martha Nussbaum, it introduces, discusses and finally analyses data with an emphasis on Martha Nussbaum’s central human functionings of practical reason and bodily health with a particular focus on the freedom to choose....

  19. TOWARD DIGITAL DYNAMIC CAPABILITIES - A THEORETICAL EXPLORATION

    Directory of Open Access Journals (Sweden)

    Zhenyu Yang

    2011-06-01

    Full Text Available Internet/Management of Information System study is a burgeoning, cross-disciplinary field in business and technology. Its development always accompanies evolving industry change, so it seems that the best strategy is not to use only existing theories for researching on-going phenomena and future evolution of Internet related issues. Alternatively, theorists in this particular field have a strong tradition of drawing from diverse aspects of social science such as human behavior, economics, strategy, and general management to derive rich discussion to address practitioners’ concerns.New theories are continuously brought into this field, particularly those that dominate in other areas with different contexts. The exposure this contributes to theory in the Information System/Internet context could potentially yield critical insights and provide feedback for further refinement of theories.This paper provides a brief review of the dynamic capability framework then incorporates the latest expansion of IT capabilities. This brings the Enterprise 2.0 system architecture into the discussion of popular management frameworks. Such an extension is expected to generate a more detailed categorization of digital dynamic capabilities. First, the key terminology of digital dynamic capabilities is presented. Second, the importance of organizational and environmental factors is emphasized in dynamic capabilities analysis. Finally, future empirical research with digital dynamic capabilities is considered.

  20. Laboratory microfusion capability study

    International Nuclear Information System (INIS)

    The purpose of this study is to elucidate the issues involved in developing a Laboratory Microfusion Capability (LMC) which is the major objective of the Inertial Confinement Fusion (ICF) program within the purview of the Department of Energy's Defense Programs. The study was initiated to support a number of DOE management needs: to provide insight for the evolution of the ICF program; to afford guidance to the ICF laboratories in planning their research and development programs; to inform Congress and others of the details and implications of the LMC; to identify criteria for selection of a concept for the Laboratory Microfusion Facility and to develop a coordinated plan for the realization of an LMC. As originally proposed, the LMC study was divided into two phases. The first phase identifies the purpose and potential utility of the LMC, the regime of its performance parameters, driver independent design issues and requirements, its development goals and requirements, and associated technical, management, staffing, environmental, and other developmental and operational issues. The second phase addresses driver-dependent issues such as specific design, range of performance capabilities, and cost. The study includes four driver options; the neodymium-glass solid state laser, the krypton fluoride excimer gas laser, the light-ion accelerator, and the heavy-ion induction linear accelerator. The results of the Phase II study are described in the present report

  1. The capability conundrum

    International Nuclear Information System (INIS)

    Full text: AMEC Nuclear recognises that a skilled workforce is essential to the success of its business. In the current global climate, the potential for new build in the UK, new build underway in Europe and across the world and the acceleration of Clean Up programmes, the competition for skilled staff is fierce. So how do companies develop and grow their Capability? What are the issues? There have been several initiatives undertaken in the UK to identify skills gaps. One example is an exercise undertaken by COGENT, the sector skills council for the Nuclear Sector, who undertook a comprehensive forward looking gap analysis for the industry in 2005 and who have now proposed a number of interventions to address the gaps identified, one of which is the creation of a Nuclear Skills Academy a second is the creation of Career Pathways. Some of the gaps identified cover shortages of skills that we face today but, more worryingly for the future, the projected down turn in the numbers of young people studying Science and Engineering subjects, both in schools and Universities and a shortage of apprenticeship places in the UK paints a gloomy picture. Where will our future capability come from? At AMEC Nuclear we recognise that we need to act now to protect our business and ensure we maintain, develop and grow our capability. We are working closely with COGENT to help to develop Career Pathways to encourage young people to the industry and we are delighted to have secured a place on the shadow board of the National Nuclear Skills Academy, an exciting project which will provide a focus for excellence in workforce skills development. Employers in the Nuclear industry have said that they want a Skills Academy that will provide national leadership in training provision and will move employers and employees from their current level of skills to those which will be needed to take these critical industries forward into a sustainable future in the UK. In order for us to assess what we

  2. Knowledge Management Capabilities Rubrics

    Directory of Open Access Journals (Sweden)

    Azizah B.A. Rahman

    2011-01-01

    Full Text Available Problem statement: Recently researchers discerned the vitality and importance of Knowledge Management Capabilities (KMC evaluation in organizations. In fact evaluation of KMC helps to prevent failure in Knowledge Management (KM projects. Approach: One of the most popular methods in the phase of evaluating KMC is Fuzzy method which evaluates seven attributes of KMC. Fuzzy needs KM experts to give their opinion about these attributes as input data. However in some organizations these experts are not available. Results: Therefore in this study a rubric matrix is developed as an assessment tool with ordered rank (very high, medium and very low of descriptive characteristics of criteria (seven attributes that organizations wish to evaluate. Conclusion: This rubric is applicable for members of an organization which are not familiar completely with KMC and also will be maintained by analyzing and surveying many different researches.

  3. PHOBOS physics capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Baker, M.D. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1995-07-15

    PHOBOS is the name of a detector and of a research program to study systematically the physics of relativistic heavy-ion collisions over a large range of impact parameters and nuclear species. Collisions with a center mass energy of 200 A GeV at RHIC are expected to produce the highest energy densities ever accessible in the laboratory. In this writeup, the authors outline the physics capabilities of the PHOBOS detector and describe the detector design in terms of the general philosophy behind the PHOBOS research program. In order to make the discussion concrete, they then focus on two specific examples of physics measurements that they plan to make at RHIC: dN/d{eta} for charged particles and the mass spectrum from {phi} {yields} K{sup +}K{sup {minus}} decays.

  4. General shape optimization capability

    Science.gov (United States)

    Chargin, Mladen K.; Raasch, Ingo; Bruns, Rudolf; Deuermeyer, Dawson

    1991-01-01

    A method is described for calculating shape sensitivities, within MSC/NASTRAN, in a simple manner without resort to external programs. The method uses natural design variables to define the shape changes in a given structure. Once the shape sensitivities are obtained, the shape optimization process is carried out in a manner similar to property optimization processes. The capability of this method is illustrated by two examples: the shape optimization of a cantilever beam with holes, loaded by a point load at the free end (with the shape of the holes and the thickness of the beam selected as the design variables), and the shape optimization of a connecting rod subjected to several different loading and boundary conditions.

  5. Incorporating Feminist Standpoint Theory

    DEFF Research Database (Denmark)

    Ahlström, Kristoffer

    2005-01-01

    As has been noted by Alvin Goldman, there are some very interesting similarities between his Veritistic Social Epistemology (VSE) and Sandra Harding’s Feminist Standpoint Theory (FST). In the present paper, it is argued that these similarities are so significant as to motivate an incorporation...

  6. Differentiating leucine incorporation of

    NARCIS (Netherlands)

    Yokokawa, T.; Sintes, E.; de Corte, D.; Olbrich, K.; Herndl, G.J.

    2012-01-01

    The abundance (based on catalyzed reporter deposition-fluorescence in situ hybrid ization, CARD-FISH) and leucine incorporation rates of Archaea and Bacteria were determined throughout the water column in the eastern Atlantic. Bacteria dominated throughout the water column, although their contributi

  7. Enhanced ocean observational capability

    Energy Technology Data Exchange (ETDEWEB)

    Volpe, A M; Esser, B K

    2000-01-10

    Coastal oceans are vital to world health and sustenance. Technology that enables new observations has always been the driver of discovery in ocean sciences. In this context, we describe the first at sea deployment and operation of an inductively coupled plasma mass spectrometer (ICPMS) for continuous measurement of trace elements in seawater. The purpose of these experiments was to demonstrate that an ICPMS could be operated in a corrosive and high vibration environment with no degradation in performance. Significant advances occurred this past year due to ship time provided by Scripps Institution of Oceanography (UCSD), as well as that funded through this project. Evaluation at sea involved performance testing and characterization of several real-time seawater analysis modes. We show that mass spectrometers can rapidly, precisely and accurately determine ultratrace metal concentrations in seawater, thus allowing high-resolution mapping of large areas of surface seawater. This analytical capability represents a significant advance toward real-time observation and understanding of water mass chemistry in dynamic coastal environments. In addition, a joint LLNL-SIO workshop was convened to define and design new technologies for ocean observation. Finally, collaborative efforts were initiated with atmospheric scientists at LLNL to identify realistic coastal ocean and river simulation models to support real-time analysis and modeling of hazardous material releases in coastal waterways.

  8. Mobile systems capability plan

    International Nuclear Information System (INIS)

    This plan was prepared to initiate contracting for and deployment of these mobile system services. 102,000 cubic meters of retrievable, contact-handled TRU waste are stored at many sites around the country. Also, an estimated 38,000 cubic meters of TRU waste will be generated in the course of waste inventory workoff and continuing DOE operations. All the defense TRU waste is destined for disposal in WIPP near Carlsbad NM. To ship TRU waste there, sites must first certify that the waste meets WIPP waste acceptance criteria. The waste must be characterized, and if not acceptable, subjected to additional processing, including repackaging. Most sites plan to use existing fixed facilities or open new ones between FY1997-2006 to perform these functions; small-quantity sites lack this capability. An alternative to fixed facilities is the use of mobile systems mounted in trailers or skids, and transported to sites. Mobile systems will be used for all characterization and certification at small sites; large sites can also use them. The Carlsbad Area Office plans to pursue a strategy of privatization of mobile system services, since this offers a number of advantages. To indicate the possible magnitude of the costs of deploying mobile systems, preliminary estimates of equipment, maintenance, and operating costs over a 10-year period were prepared and options for purchase, lease, and privatization through fixed-price contracts considered

  9. Small rover exploration capabilities

    Science.gov (United States)

    Salotti, Jean-Marc; Laithier, Corentin; Machut, Benoit; Marie, Aurélien; Bruneau, Audrey; Grömer, Gernot; Foing, Bernard H.

    2015-05-01

    For a human mission to the Moon or Mars, an important question is to determine the best strategy for the choice of surface vehicles. Recent studies suggest that the first missions to Mars will be strongly constrained and that only small unpressurized vehicles will be available. We analyze the exploration capabilities and limitations of small surface vehicles from the user perspective. Following the “human centered design” paradigm, the team focused on human systems interactions and conducted the following experiments: - Another member of our team participated in the ILEWG EuroMoonMars 2013 simulation at the Mars Desert Research Station in Utah during the same period of time. Although the possible traverses were restricted, a similar study with analog space suits and quads has been carried out. - Other experiments have been conducted in an old rock quarry close to Bordeaux, France. An expert in the use of quads for all types of terrains performed a demonstration and helped us to characterize the difficulties, the risks and advantages and drawbacks of different vehicles and tools. The vehicles that will be used on the surface of Mars have not been defined yet. Nevertheless, the results of our project already show that using a light and unpressurized vehicle (in the order of 150 kg) for the mobility on the Martian surface can be a true advantage. Part of the study was dedicated to the search for appropriate tools that could be used to make the vehicles easier to handle, safer to use and more efficient in the field to cross an obstacle. The final recommendation is to use winches and ramps, which already are widely used by quad drivers. We report on the extension of the reachable areas if such tools were available. This work has been supported by ILEWG, EuroMoonMars and the Austrian Space Forum (OEWF).

  10. Impact of Personnel Capabilities on Organizational Innovation Capability

    DEFF Research Database (Denmark)

    Nielsen, Susanne Balslev; Momeni, Mostafa

    2016-01-01

    One of the most dynamic capabilities that lead to the strongest competitive advantage in the organizations is the innovation capability. Analysing the development of a firm’s innovation capability is an important research project, and can help organizations to achieve competitive advantage in thi...

  11. The CERBERUS code: Experiments with parallel processing using RELAP5/MOD3

    International Nuclear Information System (INIS)

    CERBERUS, a six-equation parallel thermal-hydraulic system simulation code, is being developed at the Idaho National Engineering Laboratory (INEL). CERBERUS Ver.00 performs parallel computations only for the heat transfer model. It is projected that CERBERUS Ver.01 will have a parallel heat transfer and hydraulic module, excluding the matrix solver, and CERBERUS Ver.02 will contain Ver.01 plus the solver. Three implementations of the CERBERUS Ver.00 code with constructs of varying overhead have been developed using a META language. These implementations are under study on shared-memory Cray-like computer architectures. Results for the hybrid code version, which utilizes all three construct sets simultaneously (i.e., CRAY AUTO, MICRO, and MULTI TASKING) on 2- and 8-CPU Cray machines, indicate the importance of load balancing for overhead reduction, and indicate that greater speedup factors may be achievable than previously believed with a RELAP-based parallel code. Extrapolations based on Y-MP/832 overhead measurements indicate that a speedup factor of > 10 may be obtainable with the CERBERUS Ver.02 code on a 16-CPU machine

  12. Development and assessment of a modified version of RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.T. [Paul Scherrer Institute, Villigen (Switzerland)

    1995-09-01

    A summary of a number of modifications introduced in RELAP/MOD3 is presented. These include implementation of different heat transfer packages for different processes, modification of the low mass-flux Groeneveld CHF look-up table and of the dispersed flow interfacial area (and shear) as well as of the criterion for transition into and out from this regime, elimination of the under-relaxation schemes of the interfacial closure coefficients etc. The modified code is assessed against a number of separate-effect and integral test experiments and in contrast to the frozen version, is shown to result in physically sound predictions which are close to the measurements.

  13. Fukushima type severe accident analysis of Laguna Verde Nuclear Power Plant using RELAP5/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Indian Inst. of Technology, Nuclear Engineering and Technology Programme, Kanpur (India); Khanna, A., E-mail: akhanna@iitk.ac.in [Indian Inst. of Technology, Dept. of Chemical Engineering, Kanpur (India); Munshi, P. [Indian Inst. of Technology, Nuclear Engineering and Technology Programme, Kanpur (India); Allison, C. [Innovative Systems Software, Idaho Falls, Idaho (United States)

    2012-07-01

    Station Black out (SBO) in Boiling Water Reactor of Laguna Verde Nuclear Power Plant (LVNPP) is analyzed. Each flow channel is modeled as a pipe divided into 14 nodes. Physical and thermodynamic properties of both recirculation loops are identical. There are four steam lines and they are considered separately in the model. SBO lead to loss of cooling in the core, severe damage of the fuel and hydrogen production. The maximum core surface temperature has gone up to 3000 K with total hydrogen accumulation about 430 kg. The maximum debris temperature in the lower plenum is 4233 K. (author)

  14. Sensitivity analysis of the RELAP5 nodalization to IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    The main aim of this work is to identify how much the code results are affected by code user in the choice of, for example, the number of thermal-hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previous validated nodalization for analysis of steady state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of model. The results found highlight the necessity of sensitivity analysis to obtain the ideal simulation model of a system. (author)

  15. Application of the RELAP5 to the analysis of large scale integral experiments

    International Nuclear Information System (INIS)

    The present paper discusses the application of the code-modalisation to the analysis of experiments performed in the UPTF and the PANDA facility available in Germany and in Switzerland, respectively. The UPTF simulates all the internals of the reactor pressure vessel of a Pressurized Water Reactor with 1/1 scale of the geometric dimensions (diameters and lengths). The PANDA simulates the containment system of a Simplified Biling Water Reactor and the vessel of the reactor. The considered experimental dana base includes the occurrence of three-dimensional phenomena that are relevant to the refill phase of a large break Loss of Coolant Accident in a PWR and to the coupling between primary system and containment performance in a SBWR. The application of the code also required to adapt and to extend the methodology for modalisation development. The results from the qualitatively point of view are satisfactory as far as the performance of code-modalisation is concerned and do not show any important code limitation. They also confirm the suitability of the code for applications in the nuclear technology. However, this conclusion should be considered as preliminary because of the lack of independent proofs. (author)

  16. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E.; Ivanova, A. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment.

  17. Sandia Laboratories technical capabilities: testing

    International Nuclear Information System (INIS)

    The testing capabilities at Sandia Laboratories are characterized. Selected applications of these capabilities are presented to illustrate the extent to which they can be applied in research and development programs

  18. The development of capability indicators

    NARCIS (Netherlands)

    Anand, Paul; Hunter, Graham; Carter, Ian; Dowding, Keith; Guala, Francesco; Van Hees, Martin

    2009-01-01

    This paper is motivated by sustained interest in the capabilities approach to welfare economics combined with the paucity of economic statistics that measure capabilities at the individual level. Specifically, it takes a much discussed account of the normatively desirable capabilities constitutive o

  19. The Capability to Hold Property

    NARCIS (Netherlands)

    Claassen, Rutger

    2015-01-01

    This paper discusses the question of whether a capability theory of justice (such as that of Martha Nussbaum) should accept a basic “capability to hold property.” Answering this question is vital for bridging the gap between abstract capability theories of justice and their institutional implication

  20. Capability-based computer systems

    CERN Document Server

    Levy, Henry M

    2014-01-01

    Capability-Based Computer Systems focuses on computer programs and their capabilities. The text first elaborates capability- and object-based system concepts, including capability-based systems, object-based approach, and summary. The book then describes early descriptor architectures and explains the Burroughs B5000, Rice University Computer, and Basic Language Machine. The text also focuses on early capability architectures. Dennis and Van Horn's Supervisor; CAL-TSS System; MIT PDP-1 Timesharing System; and Chicago Magic Number Machine are discussed. The book then describes Plessey System 25

  1. Organizational capabilities and industry dynamics: a computational model.

    OpenAIRE

    Corsino, Marco; Gabriele, Roberto; Zaninotto, Enrico

    2008-01-01

    In this paper we propose a model of bounded rational organizations that addresses the role of organizational capabilities in shaping firm size, growth rates and profitability. Our approach aims at reconciling the logic behind stochastic models of firm growth with the notion of organizational capabilities as drivers of economic performance. We extend the stochastic framework by incorporating behavioural assumptions on: (a) the interactions between the firm and the business environment; and (b)...

  2. Capabilities and Incapabilities of the Capabilities Approach to Health Justice.

    Science.gov (United States)

    Selgelid, Michael J

    2016-01-01

    This first part of this article critiques Sridhar Venkatapuram's conception of health as a capability. It argues that Venkatapuram relies on the problematic concept of dignity, implies that those who are unhealthy lack lives worthy of dignity (which seems politically incorrect), sets a low bar for health, appeals to metaphysically problematic thresholds, fails to draw clear connections between appealed-to capabilities and health, and downplays the importance/relevance of health functioning. It concludes by questioning whether justice entitlements should pertain to the capability for health versus health achievements, challenging Venkatapuram's claims about the strength of health entitlements, and demonstrating that the capabilities approach is unnecessary to address social determinants of health.

  3. Transactional capability: Innovation's missing link

    Directory of Open Access Journals (Sweden)

    Jorge Tello-Gamarra

    2013-06-01

    Full Text Available The topic of innovation as a source of competitive advantage for firms is consolidated in the literature. An innovation-based advantage is generally achieved by following a structured process, in which technological capability plays a key role. However, a question remains in studies into innovation, which is concerned with the reasons why not all firms that develop a technological capability are innovative. Where is the missing link? Great efforts have been made in attempts to answer to this question. However, the study of this link as a source of advantage for firms deserves further attention. This article aims to present a framework with two essential dimensions: (1 the technological capability and (2 the transactional capability. Technological capability is the ability of firms to make effective use of technical knowledge in order to improve production processes and develop new products and services. Transactional capability is defined as a repertoire of abilities, processes, experiences, skills, knowledge and routines that the firm uses to minimize its transaction costs (ex¿ante and ex¿post. Given that the present study is a theoretical paper, methodologically it is based on a literature review. The main finding of this study is the identification of transactional capability as the missing link in innovation. Accordingly, to be innovative, in addition to having a technological capability, a firm needs to develop its transactional capability.

  4. Technological Dynamics and Social Capability

    DEFF Research Database (Denmark)

    Fagerberg, Jan; Feldman, Maryann; Srholec, Martin

    2014-01-01

    for the sample as a whole between 1998 and 2008. The results indicate that social capabilities, such as well-developed public knowledge infrastructure, an egalitarian distribution of income, a participatory democracy and prevalence of public safety condition the growth of technological capabilities. Possible......This article analyzes factors shaping technological capabilities in USA and European countries, and shows that the differences between the two continents in this respect are much smaller than commonly assumed. The analysis demonstrates a tendency toward convergence in technological capabilities...

  5. Expanding capabilities of the debris analysis workstation

    Science.gov (United States)

    Spencer, David B.; Sorge, Marlon E.; Mains, Deanna L.; Shubert, Ann J.; Gerhart, Charlotte M.; Yates, Ken W.; Leake, Michael

    1996-10-01

    Determining the hazards from debris-generating events is a design and safety consideration for a number of space systems, both currently operating and planned. To meet these and other requirements, the United States Air Force (USAF) Phillips Laboratory (PL) Space Debris Research Program has developed a simulation software package called the Debris Analysis Workstation (DAW). This software provides an analysis capability for assessing a wide variety of debris hazards. DAW integrates several component debris analysis models and data visualization tools into a single analysis platform that meets the needs for Department of Defense space debris analysis, and is both user friendly and modular. This allows for studies to be performed expeditiously by analysts who are not debris experts. The current version of DAW includes models for spacecraft breakup, debris orbital lifetime, collision hazard risk assessment, and collision dispersion, as well as a satellite catalog database manager, a drag inclusive propagator, a graphical user interface, and data visualization routines. Together they provide capabilities to conduct several types of analyses, ranging from range safety assessments to satellite constellation risk assessment. Work is progressing to add new capabilities with the incorporation of additional models and improved designs. The existing tools are in their initial integrated form, but the 'glue' that will ultimately bring them together into an integrated system is an object oriented language layer scheduled to be added soon. Other candidate component models under consideration for incorporation include additional orbital propagators, error estimation routines, other dispersion models, and other breakup models. At present, DAW resides on a SUNR workstation, although future versions could be tailored for other platforms, depending on the need.

  6. Emissive sensors and devices incorporating these sensors

    Science.gov (United States)

    Swager, Timothy M; Zhang, Shi-Wei

    2013-02-05

    The present invention generally relates to luminescent and/or optically absorbing compositions and/or precursors to those compositions, including solid films incorporating these compositions/precursors, exhibiting increased luminescent lifetimes, quantum yields, enhanced stabilities and/or amplified emissions. The present invention also relates to sensors and methods for sensing analytes through luminescent and/or optically absorbing properties of these compositions and/or precursors. Examples of analytes detectable by the invention include electrophiles, alkylating agents, thionyl halides, and phosphate ester groups including phosphoryl halides, cyanides and thioates such as those found in certain chemical warfare agents. The present invention additionally relates to devices and methods for amplifying emissions, such as those produced using the above-described compositions and/or precursors, by incorporating the composition and/or precursor within a polymer having an energy migration pathway. In some cases, the compositions and/or precursors thereof include a compound capable of undergoing a cyclization reaction.

  7. Ionic liquid incorporating thiosalicylate for metal removal

    Science.gov (United States)

    Wilfred, Cecilia Devi; Mustafa, Fadwa Babiker; Romeli, Fatimah Julia

    2012-09-01

    Ionic liquids are a class of organic molten salts "designer solvents" that are composed totally of anions (inorganic and organic polyatomic) and organic cations. The replacement of volatile organic solvents from a separation process is of utmost importance since the use of a large excess of these solvents is hazardous and creates ecological problem. The new method for metal ion extraction is by using task-specific ionic liquids such as ionic liquids which incorporate thiosalicylate functionality. This paper looks at producing a new cluster of ionic liquids which incorporates thiosalicylate with pyridinium cation. Its thermophysical properties such as density and viscosity in single and binary mixtures are studied. The ionic liquids' capability in metal removal processes is evaluated.

  8. Identifying 21st Century Capabilities

    Science.gov (United States)

    Stevens, Robert

    2012-01-01

    What are the capabilities necessary to meet 21st century challenges? Much of the literature on 21st century skills focuses on skills necessary to meet those challenges associated with future work in a globalised world. The result is a limited characterisation of those capabilities necessary to address 21st century social, health and particularly…

  9. A Multivariate Process Capability Index

    Institute of Scientific and Technical Information of China (English)

    WANG Shaoxi; JIA Xinzhang; JIAO Huifang; BIAN Yingchao; SONG Ning; ZHAO Luyu; WEN Xin

    2006-01-01

    Process capability indices have been widely used in the manufacturing industry, providing numerical measures on process precision, process accuracy, and process performance. Capability measures for processes with a single characteristic have been investigated extensively. However, capability measures for processes with multiple characteristics are comparatively neglected. In this paper, inspired by the approach and model of process capability index investigated by K.S.Chen et al.(2003) and A.B.Yeh et al.(1998), a note model of multivariate process capability index based on non-conformity is presented. As for this index, the data of each single characteristic don't require satisfying normal distribution, of which its computing is simple and particioners will not fell too theoretical. At last the application analysis is made.

  10. Developing Collaborative Product Development Capabilities

    DEFF Research Database (Denmark)

    Mahnke, Volker; Tran, Yen

    2012-01-01

    innovation strategies’. Our analyses suggest that developing such collaboration capabilities benefits from the search for complementary practices, the combination of learning styles, and the development of weak and strong ties. Results also underscore the crucial importance of co-evolution of multi......Collaborative product development capabilities support a company’s product innovation activities. In the context of the fast fashion sector, this paper examines the development of the product development capabilities (PDC) that align product development capabilities in a dual innovation context......, one, slow paced, where the firm is well established and the other, fast paced, which represents a new competitive arena in which the company competes. To understand the process associated with collaborative capability development, we studied three Scandinavian fashion companies pursuing ‘dual...

  11. Nepal CRS project incorporates.

    Science.gov (United States)

    1983-01-01

    The Nepal Contraceptive Retail Sales (CRS) Project, 5 years after lauching product sales in June 1978, incorporated as a private, nonprofit company under Nepalese management. The transition was finalized in August 1983. The Company will work through a cooperative agreement with USAID/Kathmandu to complement the national family planning goals as the program continues to provide comtraceptives through retail channels at subsidized prices. Company objectives include: increase contraceptive sales by at least 15% per year; make CRS cost effective and move towards self sufficiency; and explore the possibility of marketing noncontraceptive health products to improve primary health care. After only5 years the program can point to some impressive successes. The number of retial shops selling family planning products increased from 100 in 1978 to over 8000, extending CRS product availability to 66 of the country's 75 districts. Retail sales have climbed dramatically in the 5-year period, from Rs 46,817 in 1978 to Rs 271,039 in 1982. Sales in terms of couple year protection CYP) have grown to 24,451 CYP(1982), a 36% increase over 1980 CYP. Since the beginning of the CRS marketing program, total distribution of contraceptives--through both CRS and the Family Planning Maternal and Child Haelth (FP/MCH) Project--has been increasing. While the FP/MCH program remains the largest distributor,contribution of CRS Products is increasing, indicating that CRS is creating new product acceptors. CRS market share in 1982 was 43% for condoms and 16% for oral contraceptives (OCs). CRS markets 5 products which are subsidized in order to be affordable to consumers as well as attractive to sellers. The initial products launched in June 1978 were Gulaf standard dose OCs and Dhaal lubricated colored condoms. A less expensive lubricates, plain Suki-Dhaal condom was introduced in June 1980 in an attempt to reach poorer rural populations, but rural distribution costs are excessive and Suki

  12. Nepal CRS project incorporates.

    Science.gov (United States)

    1983-01-01

    The Nepal Contraceptive Retail Sales (CRS) Project, 5 years after lauching product sales in June 1978, incorporated as a private, nonprofit company under Nepalese management. The transition was finalized in August 1983. The Company will work through a cooperative agreement with USAID/Kathmandu to complement the national family planning goals as the program continues to provide comtraceptives through retail channels at subsidized prices. Company objectives include: increase contraceptive sales by at least 15% per year; make CRS cost effective and move towards self sufficiency; and explore the possibility of marketing noncontraceptive health products to improve primary health care. After only5 years the program can point to some impressive successes. The number of retial shops selling family planning products increased from 100 in 1978 to over 8000, extending CRS product availability to 66 of the country's 75 districts. Retail sales have climbed dramatically in the 5-year period, from Rs 46,817 in 1978 to Rs 271,039 in 1982. Sales in terms of couple year protection CYP) have grown to 24,451 CYP(1982), a 36% increase over 1980 CYP. Since the beginning of the CRS marketing program, total distribution of contraceptives--through both CRS and the Family Planning Maternal and Child Haelth (FP/MCH) Project--has been increasing. While the FP/MCH program remains the largest distributor,contribution of CRS Products is increasing, indicating that CRS is creating new product acceptors. CRS market share in 1982 was 43% for condoms and 16% for oral contraceptives (OCs). CRS markets 5 products which are subsidized in order to be affordable to consumers as well as attractive to sellers. The initial products launched in June 1978 were Gulaf standard dose OCs and Dhaal lubricated colored condoms. A less expensive lubricates, plain Suki-Dhaal condom was introduced in June 1980 in an attempt to reach poorer rural populations, but rural distribution costs are excessive and Suki

  13. Technological capability at the Brazilian official pharmaceutical laboratories

    Directory of Open Access Journals (Sweden)

    José Vitor Bomtempo Martins

    2008-10-01

    Full Text Available This paper studies the technological capability in the Brazilian Official Pharmaceutical Laboratories [OPL]. The technological capability analysis could contribute to organization strategies and governmental actions in order to improve OPL basic tasks as well to incorporate new ones, particularly concerning the innovation management. Inspired in Figueiredo (2000, 2003a, 2003b and Figueiredo and Ariffin (2003, a framework was drawn and adapted to pharmaceutical industry characteristics and current sanitary and health legislation. The framework allows to map different dimensions of the technological capability (installations, processes, products, equipments, organizational capability and knowledge management and the level attained by OPL (ordinary or innovating capability. OPL show a good development of ordinary capabilities, particularly in Product and Processes. Concerning the other dimensions, OPL are quite diverse. In general, innovating capabilities are not much developed. In the short term, it was identified a dispersion in the capacitating efforts. Considering their present level and the absorption efforts, good perspectives can be found in Installations, Processes and Organizational Capability. A lower level of efforts in Products and Knowledge Management could undermine these capabilities in the future.

  14. Methodology for the Incorporation of Passive Component Aging Modeling into the RAVEN/ RELAP-7 Environment

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua; Alfonsi, Andrea; Askin Guler; Tunc Aldemir

    2014-11-01

    Passive system, structure and components (SSCs) will degrade over their operation life and this degradation may cause to reduction in the safety margins of a nuclear power plant. In traditional probabilistic risk assessment (PRA) using the event-tree/fault-tree methodology, passive SSC failure rates are generally based on generic plant failure data and the true state of a specific plant is not reflected realistically. To address aging effects of passive SSCs in the traditional PRA methodology [1] does consider physics based models that account for the operating conditions in the plant, however, [1] does not include effects of surveillance/inspection. This paper represents an overall methodology for the incorporation of aging modeling of passive components into the RAVEN/RELAP-7 environment which provides a framework for performing dynamic PRA. Dynamic PRA allows consideration of both epistemic and aleatory uncertainties (including those associated with maintenance activities) in a consistent phenomenological and probabilistic framework and is often needed when there is complex process/hardware/software/firmware/ human interaction [2]. Dynamic PRA has gained attention recently due to difficulties in the traditional PRA modeling of aging effects of passive components using physics based models and also in the modeling of digital instrumentation and control systems. RAVEN (Reactor Analysis and Virtual control Environment) [3] is a software package under development at the Idaho National Laboratory (INL) as an online control logic driver and post-processing tool. It is coupled to the plant transient code RELAP-7 (Reactor Excursion and Leak Analysis Program) also currently under development at INL [3], as well as RELAP 5 [4]. The overall methodology aims to: • Address multiple aging mechanisms involving large number of components in a computational feasible manner where sequencing of events is conditioned on the physical conditions predicted in a simulation

  15. Transforming organizational capabilities in strategizing

    DEFF Research Database (Denmark)

    Jørgensen, Claus; Friis, Ole Uhrskov; Koch, Christian

    2014-01-01

    reallocated over time thereby creating a growing need for new capabilities and transformed knowledge handling routines. IT emerged into an important resource to support more complex routines of product development as well as specific management and HRM processes assisting the transformation......Offshored and networked enterprises are becoming an important if not leading organizational form and this development seriously challenges their organizational capabilities. More specifically, over the last years, SMEs have commenced entering these kinds of arrangements. As the organizational...

  16. Earth Science Capability Demonstration Project

    Science.gov (United States)

    Cobleigh, Brent

    2006-01-01

    A viewgraph presentation reviewing the Earth Science Capability Demonstration Project is shown. The contents include: 1) ESCD Project; 2) Available Flight Assets; 3) Ikhana Procurement; 4) GCS Layout; 5) Baseline Predator B Architecture; 6) Ikhana Architecture; 7) UAV Capability Assessment; 8) The Big Picture; 9) NASA/NOAA UAV Demo (5/05 to 9/05); 10) NASA/USFS Western States Fire Mission (8/06); and 11) Suborbital Telepresence.

  17. QALYs and the capability approach.

    Science.gov (United States)

    Cookson, Richard

    2005-08-01

    This explores the applicability of Sen's capability approach to the economic evaluation of health care programmes. An individual's 'capability set' describes his freedom to choose valuable activities and states of being ('functionings'). Direct estimation and valuation of capability sets is not feasible at present. Standard preference-based methods such as willingness to pay are feasible, but problematic due to the adaptive and constructed nature of individual preferences over time and under uncertainty. An alternative is to re-interpret the QALY as a cardinal and interpersonally comparable index of the value of the individual's capability set. This approach has limitations, since the link between QALYs and capabilities is not straightforward. Nevertheless, the QALY approach is recognisable as an application of the capability approach since it pays close attention to functionings, through the use of survey-based multi-attribute health state valuation instruments, and permits conceptions of value other than the traditional utilitarian ones of choice, desire-fulfilment and happiness. Furthermore, suitably re-interpreted, it can account for (i) non-separability between health and non-health components of value; and suitably modified it can also account for (ii) process attributes of care, which may have a direct effect on non-health functionings such as comfort and dignity, and (iii) sub-group diversity in the value of the same health functionings. PMID:15693028

  18. Incorporation, plurality, and the incorporation of plurals: a dynamic approach

    NARCIS (Netherlands)

    de Swart, H.E.; Farkas, D. F.

    2004-01-01

    This paper deals with the semantic properties of incorporated nominals that are present at clausal syntax. Such nominals exhibit a complex cluster of semantic properties, ranging from argument structure, scope, and number to discourse transparency. We develop an analysis of incorporation in the fram

  19. Accelerator and electrodynamics capability review

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Kevin W [Los Alamos National Laboratory

    2010-01-01

    Los Alamos National Laboratory (LANL) uses capability reviews to assess the science, technology and engineering (STE) quality and institutional integration and to advise Laboratory Management on the current and future health of the STE. Capability reviews address the STE integration that LANL uses to meet mission requirements. The Capability Review Committees serve a dual role of providing assessment of the Laboratory's technical contributions and integration towards its missions and providing advice to Laboratory Management. The assessments and advice are documented in reports prepared by the Capability Review Committees that are delivered to the Director and to the Principal Associate Director for Science, Technology and Engineering (PADSTE). Laboratory Management will use this report for STE assessment and planning. LANL has defined fifteen STE capabilities. Electrodynamics and Accelerators is one of the seven STE capabilities that LANL Management (Director, PADSTE, technical Associate Directors) has identified for review in Fiscal Year (FY) 2010. Accelerators and electrodynamics at LANL comprise a blend of large-scale facilities and innovative small-scale research with a growing focus on national security applications. This review is organized into five topical areas: (1) Free Electron Lasers; (2) Linear Accelerator Science and Technology; (3) Advanced Electromagnetics; (4) Next Generation Accelerator Concepts; and (5) National Security Accelerator Applications. The focus is on innovative technology with an emphasis on applications relevant to Laboratory mission. The role of Laboratory Directed Research and Development (LDRD) in support of accelerators/electrodynamics will be discussed. The review provides an opportunity for interaction with early career staff. Program sponsors and customers will provide their input on the value of the accelerator and electrodynamics capability to the Laboratory mission.

  20. Incorporation of NREL Solar Advisor Model Photovoltaic Capabilities with GridLAB-D

    Energy Technology Data Exchange (ETDEWEB)

    Tuffner, Francis K.; Hammerstrom, Janelle L.; Singh, Ruchi

    2012-10-19

    This report provides a summary of the work updating the photovoltaic model inside GridLAB-D. The National Renewable Energy Laboratory Solar Advisor Model (SAM) was utilized as a basis for algorithms and validation of the new implementation. Subsequent testing revealed that the two implementations are nearly identical in both solar impacts and power output levels. This synergized model aides the system-level impact studies of GridLAB-D, but also allows more specific details of a particular site to be explored via the SAM software.

  1. GCN capabilities and status, and the incorporation of LIGO/Virgo

    Science.gov (United States)

    Barthelmy, Scott

    2016-03-01

    The Gamma-ray Coordinates Network / Transient Astronomy Network (GCN/TAN) is a single-point source for all transient astronomy notification. It collects the astrophysical transients from the missions (space-based and nearly all ground-based), puts them into a standard format, and distributes them to whomever wants to receive them. This is all done autonomously (completely autonomous within GCN/TAN, and almost always autonomously within the producer end of operations). This automation means minimal time delays (Collaboration (LVC) Notices are now implemented in the GCN/TAN system. During the proprietary phase, the recipients must have an MoU with LVC and LVC must authorize GCN to distribute LVC Notices to each given MoU follow-up observer. In addition to Notices, there are the GCN Circulars, which are prose-style reports of follow-up observations made and results obtains. During the LVC Proprietary phase there are also the GCN LVC Circulars, which also require authorization from LVC to join the LVC Circulars.

  2. Benefits of oxygen incorporation in atomic laminates

    Science.gov (United States)

    Dahlqvist, Martin

    2016-04-01

    Atomic laminates such as MAX phases benefit from the addition of oxygen in many ways, from the formation of a protective oxide surface layer with self-healing capabilities when cracks form to the tuning of anisotropic conductivity. In this paper oxygen incorporation and vacancy formation in M 2AlC (M  =  Ti, V, Cr) MAX phases have been studied using first-principles calculations where the focus is on phase stability and electronic structure for different oxygen and/or vacancy configurations. Oxygen prefers different lattice sites depending on M-element and this can be correlated to the number of available non-bonding M d-electrons. In Ti2AlC, oxygen substitutes carbon while in Cr2AlC it is located interstitially within the Al-layer. I predict that oxygen incorporation in Ti2AlC stabilizes the material, which explains the experimentally observed 12.5 at% oxygen (x  =  0.5) in Ti2Al(C1‑x O x ). In addition, it is also possible to use oxygen to stabilize the hypothetical Zr2AlC and Hf2AlC. Hence, oxygen incorporation may be beneficial in many ways. Not only can it make a material more stable, but it also can act as a reservoir for internal self-healing with shorter diffusion paths.

  3. Benefits of oxygen incorporation in atomic laminates

    International Nuclear Information System (INIS)

    Atomic laminates such as MAX phases benefit from the addition of oxygen in many ways, from the formation of a protective oxide surface layer with self-healing capabilities when cracks form to the tuning of anisotropic conductivity. In this paper oxygen incorporation and vacancy formation in M 2AlC (M  =  Ti, V, Cr) MAX phases have been studied using first-principles calculations where the focus is on phase stability and electronic structure for different oxygen and/or vacancy configurations. Oxygen prefers different lattice sites depending on M-element and this can be correlated to the number of available non-bonding M d-electrons. In Ti2AlC, oxygen substitutes carbon while in Cr2AlC it is located interstitially within the Al-layer. I predict that oxygen incorporation in Ti2AlC stabilizes the material, which explains the experimentally observed 12.5 at% oxygen (x  =  0.5) in Ti2Al(C1−xOx). In addition, it is also possible to use oxygen to stabilize the hypothetical Zr2AlC and Hf2AlC. Hence, oxygen incorporation may be beneficial in many ways. Not only can it make a material more stable, but it also can act as a reservoir for internal self-healing with shorter diffusion paths. (paper)

  4. A unifying process capability metric

    Directory of Open Access Journals (Sweden)

    John Jay Flaig

    2009-07-01

    Full Text Available A new economic approach to process capability assessment is presented, which differs from the commonly used engineering metrics. The proposed metric consists of two economic capability measures – the expected profit and the variation in profit of the process. This dual economic metric offers a number of significant advantages over other engineering or economic metrics used in process capability analysis. First, it is easy to understand and communicate. Second, it is based on a measure of total system performance. Third, it unifies the fraction nonconforming approach and the expected loss approach. Fourth, it reflects the underlying interest of management in knowing the expected financial performance of a process and its potential variation.

  5. Linking strategic flexibility and operational efficiency: the mediating role of ambidextrous operational capabilities

    NARCIS (Netherlands)

    S. Kortmann; C. Gelhard; C. Zimmermann; F.T. Piller

    2014-01-01

    We elucidate the important, though complex, relationship between strategic flexibility and operational efficiency. We incorporate insights from the dynamic resource-based view, ambidexterity literature and managerial practice to explain how two ambidextrous operational capabilities, i.e., mass custo

  6. NWCF maintenance features and capabilities

    International Nuclear Information System (INIS)

    A New Waste Calcining Facility is being built at the Idaho Chemical Processing Plant to replace the existing Waste Calcining Facility which was built to demonstrate fluidized-bed solidification of highly radioactive liquid wastes. The new facility is being designed to provide a higher waste throughput, more corrosion resistant materials of construction, more effective cleanup of effluent streams, and extensive remote maintenance and equipment replacement capability. The facility will also contain extensive decontamination capability should contact maintenance become necessary. The facility is presently in construction and is scheduled for hot operation in 1980

  7. Judgmental Forecasting of Operational Capabilities

    DEFF Research Database (Denmark)

    Hallin, Carina Antonia; Tveterås, Sigbjørn; Andersen, Torben Juul

    This paper explores a new judgmental forecasting indicator, the Employee Sensed Operational Capabilities (ESOC). The purpose of the ESOC is to establish a practical prediction tool that can provide early signals about changes in financial performance by gauging frontline employees’ sensing...

  8. Building server capabilities in China

    DEFF Research Database (Denmark)

    Adeyemi, Oluseyi; Slepniov, Dmitrij; Wæhrens, Brian Vejrum;

    2012-01-01

    The purpose of this paper is to further our understanding of multinational companies building server capabilities in China. The paper is based on the cases of two western companies with operations in China. The findings highlight a number of common patterns in the 1) managerial challenges related...

  9. Microfoundations of Routines and Capabilities

    DEFF Research Database (Denmark)

    Felin, Teppo; Foss, Nicolai Juul; Heimriks, Koen H.;

    We discuss the microfoundations of routines and capabilities, including why a microfoundations view is needed and how it may inform work on organizational and competitive heterogeneity. Building on extant research, we identify three primary categories of micro-level components underlying routines...

  10. Microfoundations of Routines and Capabilities

    DEFF Research Database (Denmark)

    Felin, Tippo; Foss, Nicolai Juul; Heimericks, Koen H.;

    2012-01-01

    This article introduces the Special Issue and discusses the microfoundations of routines and capabilities, including why a microfoundations view is needed and how it may inform work on organizational and competitive heterogeneity. Building on extant research, we identify three primary categories ...

  11. ABOUT SOLIDWORKS SUSTAINABILITY MODULE CAPABILITIES

    Directory of Open Access Journals (Sweden)

    Cătălin IANCU

    2014-05-01

    Full Text Available In the paper are presented the SolidWorks analysis steps to be taken in order to study sustainability of parts or assemblies designed. There are presented the software capabilities and the settings that have to be done for such analysis and the results shown by software.

  12. Demonstration MTI/SAR capability

    NARCIS (Netherlands)

    Vries, F.P.P. de; Broek, A.C. van den; Otten, M.P.G.; Groot, J.S.; Steeghs, T.P.H.; Dekker, R.J.; Rossum, W.L. van

    2001-01-01

    The aim of this project is to demonstrate to the Dutch armed forces the capability of MTI (Moving Target Indicator) and SAR (Synthetic Aperture Radar). This is done with the Dutch PHARUS sensor. The sensor is used to demonstrate how a phased array antenna can be used as an MTI/SAR sensor combination

  13. Synthesis and Enzymatic Incorporation of Modified Deoxyuridine Triphosphates

    Directory of Open Access Journals (Sweden)

    Erkai Liu

    2015-07-01

    Full Text Available To expand the chemical functionality of DNAzymes and aptamers, several new modified deoxyuridine triphosphates have been synthesized. An important precursor that enables this aim is 5-aminomethyl dUTP, whereby the pendent amine serves as a handle for further synthetic functionalization. Five functional groups were conjugated to 5-aminomethyl dUTP. Incorporation assays were performed on several templates that demand 2–5 sequential incorporation events using several commercially available DNA polymerases. It was found that Vent (exo- DNA polymerase efficiently incorporates all five modified dUTPs. In addition, all nucleoside triphosphates were capable of supporting a double-stranded exponential PCR amplification. Modified PCR amplicons were PCR amplified into unmodified DNA and sequenced to verify that genetic information was conserved through incorporation, amplification, and reamplification. Overall these modified dUTPs represent new candidate substrates for use in selections using modified nucleotide libraries.

  14. Human-Centered Design Capability

    Science.gov (United States)

    Fitts, David J.; Howard, Robert

    2009-01-01

    For NASA, human-centered design (HCD) seeks opportunities to mitigate the challenges of living and working in space in order to enhance human productivity and well-being. Direct design participation during the development stage is difficult, however, during project formulation, a HCD approach can lead to better more cost-effective products. HCD can also help a program enter the development stage with a clear vision for product acquisition. HCD tools for clarifying design intent are listed. To infuse HCD into the spaceflight lifecycle the Space and Life Sciences Directorate developed the Habitability Design Center. The Center has collaborated successfully with program and project design teams and with JSC's Engineering Directorate. This presentation discusses HCD capabilities and depicts the Center's design examples and capabilities.

  15. Improving the RPC rate capability

    CERN Document Server

    Aielli, G; Cardarelli, R; Di Ciaccio, A; Di Stante, L; Iuppa, R; Liberti, B; Paolozzi, L; Pastori, E; Santonico, R; Toppi, M

    2016-01-01

    This paper has the purpose to study the rate capability of the Resistive Plate Chamber, RPC, starting from the basic physics of this detector. The effect of different working parameters determining the rate capability is analysed in detail, in order to optimize a new family of RPCs for applications to heavy irradiation environments and in particular to the LHC phase 2. A special emphasis is given to the improvement achievable by minimizing the avalanche charge delivered in the gas. The paper shows experimental results of Cosmic Ray tests, performed to study the avalanche features for different gas gap sizes, with particular attention to the overall delivered charge. For this purpose, the paper studies, in parallel to the prompt electronic signal, also the ionic signal which gives the main contribution to the delivered charge. Whenever possible the test results are interpreted on the base of the RPC detector physics and are intended to extend and reinforce our physical understanding of this detector.

  16. Active containment systems incorporating modified pillared clays

    Energy Technology Data Exchange (ETDEWEB)

    Lundie, P. [Envirotech (Scotland) Ltd., Aberdeen (United Kingdom)]|[Environmental Resource Industries Disposal Pty Ltd., Perth (Australia); McLeod, N. [Envirotreat Ltd., Kingswinford (United Kingdom)

    1997-12-31

    The application of treatment technologies in active containment systems provides a more advanced and effective method for the remediation of contaminated sites. These treatment technologies can be applied in permeable reactive walls and/or funnel and gate systems. The application of modified pillared clays in active containment systems provides a mechanism for producing permeable reactive walls with versatile properties. These pillared clays are suitably modified to incorporate reactive intercalatants capable of reacting with both a broad range of organic pollutants of varying molecular size, polarity and reactivity. Heavy metals can be removed from contaminated water by conventional ion-exchange and other reactive processes within the clay structure. Complex contamination problems can be addressed by the application of more than one modified clay on a site specific basis. This paper briefly describes the active containment system and the structure/chemistry of the modified pillared clay technology, illustrating potential applications of the in-situ treatment process for contaminated site remediation.

  17. RELAP-7 Beta Release: Summary of Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Martineau, Richard C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-12-01

    RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory (INL). Building upon the decades of software development at the INL, we began the development of RELAP-7 in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway. As part of this development, the first lines of RELAP-7 code were committed to the software revision control repository on November 7th, 2011. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. RELAP-7 is built using the INL’s modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved numerical calculations (including higher-order integration in both space and time, yielding converged second-order accuracy). The RELAP-7 code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. Each component can have options for different fluid models such as single- and two-phase flow. This component-based and physics-based software architecture allows RELAP-7 to adopt different physical models for different applications. A relatively new two-phase hydrodynamic model, termed the ''7-Equation model'' for two phasic pressures, velocities, energies, and volumetric fraction, is incorporated into RELAP-7 for liquid-gas (water-steam) flows. This new model allows second-order integration because it is well-posed, which will reduce the numerical error associated with traditional systems analysis codes. In this paper, we provide a RELAP-7 capability list describing analysis features, range of applicability, and reactor components that will be available for the December 15th, 2014 beta release of the software.

  18. Determining your organization's 'risk capability'.

    Science.gov (United States)

    Hannah, Bill; Hancock, Melinda

    2014-05-01

    An assessment of a provider's level of risk capability should focus on three key elements: Business intelligence, including sophisticated analytical models that can offer insight into the expected cost and quality of care for a given population. Clinical enterprise maturity, marked by the ability to improve health outcomes and to manage utilization and costs to drive change. Revenue transformation, emphasizing the need for a revenue cycle platform that allows for risk acceptance and management and that provides incentives for performance against defined objectives. PMID:24851456

  19. Developing a dispersant spraying capability

    Energy Technology Data Exchange (ETDEWEB)

    Gill, S.D.

    1979-01-01

    In developing a national dispersant spraying capability, the Canadian Coast Guard (CCG) has undertaken a modification program to enable the conventional offshore spraying gear to be mounted on almost any vessel of convenience. Smaller, more versatile inshore spraying vessels and pumps have been designed and built. With the popularization of concentrated dispersants, the inshore pumping equipment can be used aboard hovercraft for special application situations. A program of acquiring mobile dispersant storage tanks has been undertaken with auxiliary equipment that will facilitate the shipment of dispersants in bulk by air freight. Work also has commenced on extending the dispersant application program to include the CCG fleet of helicopters.

  20. Production, innovation and service capabilities

    DEFF Research Database (Denmark)

    Slepniov, Dmitrij; Wæhrens, Brian Vejrum; Wu, Dong;

    2012-01-01

    Fragmentation and global dispersion are among the most prominent characteristics of contemporary operations. Not only routine transactional tasks, but also more knowledge-intensive proprietary tasks are subjected to this trend. As a result of this, complex configurations of assets and capabilities...... crossing both national and organisational borders emerge. The challenge of coordination in these configurations is an imperative which has not been adequately addressed so far. Therefore, by using explorative cases of Chinese and Danish companies, this paper seeks to develop a conceptual framework relating...

  1. Developing A/E Capabilities

    International Nuclear Information System (INIS)

    During the last few years, the methods used by EMPRESARIOS AGRUPADOS and INITEC to perform Architect-Engineering work in Spain for nuclear projects has undergone a process of significant change in project management and engineering approaches. Specific practical examples of management techniques and design practices which represent a good record of results will be discussed. They are identified as areas of special interest in developing A/E capabilities for nuclear projects . Command of these areas should produce major payoffs in local participation and contribute to achieving real nuclear engineering capabities in the country. (author)

  2. ITER EDA design confinement capability

    Science.gov (United States)

    Uckan, N. A.

    Major device parameters for ITER-EDA and CDA are given in this paper. Ignition capability of the EDA (and CDA) operational scenarios is evaluated using both the 1 1/2-D time-dependent transport simulations and 0-D global models under different confinement ((chi((gradient)(T)(sub e)(sub crit)), empirical global energy confinement scalings, chi(empirical), etc.) assumptions. Results from some of these transport simulations and confinement assessments are summarized in and compared with the ITER CDA results.

  3. ITER EDA design confinement capability

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, N.A.

    1993-06-01

    Major device parameters for ITER-EDA and CDA are given in this paper. Ignition capability of the EDA (and CDA) operational scenarios is evaluated using both the 1-1/2-D time-dependent transport simulations and 0-D global models under different confinement [{chi}({triangledown}T{sub e}){sub crit}, empirical global energy confinement scalings, {chi}(empirical), etc.] assumptions. Results from some of these transport simulations and confinement assessments are summarized in and compared with the ITER CDA reference ignition scenario.

  4. ITER EDA design confinement capability

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, N.A.

    1993-01-01

    Major device parameters for ITER-EDA and CDA are given in this paper. Ignition capability of the EDA (and CDA) operational scenarios is evaluated using both the 1-1/2-D time-dependent transport simulations and 0-D global models under different confinement [[chi]([triangledown]T[sub e])[sub crit], empirical global energy confinement scalings, [chi](empirical), etc.] assumptions. Results from some of these transport simulations and confinement assessments are summarized in and compared with the ITER CDA reference ignition scenario.

  5. Incorporation of salinity in Water Availability Modeling

    Science.gov (United States)

    Wurbs, Ralph A.; Lee, Chihun

    2011-10-01

    SummaryNatural salt pollution from geologic formations in the upper watersheds of several large river basins in the Southwestern United States severely constrains the use of otherwise available major water supply sources. The Water Rights Analysis Package modeling system has been routinely applied in Texas since the late 1990s in regional and statewide planning studies and administration of the state's water rights permit system, but without consideration of water quality. The modeling system was recently expanded to incorporate salinity considerations in assessments of river/reservoir system capabilities for supplying water for environmental, municipal, agricultural, and industrial needs. Salinity loads and concentrations are tracked through systems of river reaches and reservoirs to develop concentration frequency statistics that augment flow frequency and water supply reliability metrics at pertinent locations for alternative water management strategies. Flexible generalized capabilities are developed for using limited observed salinity data to model highly variable concentrations imposed upon complex river regulation infrastructure and institutional water allocation/management practices.

  6. Incorporating Argumentation through Forensic Science

    Science.gov (United States)

    Wheeler, Lindsay B.; Maeng, Jennifer L.; Smetana, Lara K.

    2014-01-01

    This article outlines how to incorporate argumentation into a forensic science unit using a mock trial. Practical details of the mock trial include: (1) a method of scaffolding students' development of their argument for the trial, (2) a clearly outlined set of expectations for students during the planning and implementation of the mock…

  7. Professionals and Public Good Capabilities

    Directory of Open Access Journals (Sweden)

    Melanie Walker

    2015-12-01

    Full Text Available Martha Nussbaum (2011 reminds us that, all over the world people are struggling for a life that is fully human - a life worthy of human dignity. Purely income-based and preference-based evaluations, as Sen (1999 argues, do not adequately capture what it means for each person to have quality of life. There are other things that make life good for a person, including access to publicly provided professional services. The question then is what version of education inflects more towards the intrinsic and transformational possibilities of professional work and contributions to decent societies? This paper suggests that we need a normative approach to professional education and professionalism; it is not the case that any old version will do. We also need normative criteria to move beyond social critique and to overcome a merely defensive attitude and to give a positive definition to the potential achievements of the professions. Moreover universities are connected to society, most especially through the professionals they educate; it is reasonable in our contemporary world to educate professional graduates to be in a position to alleviate inequalities, and to have the knowledge, skills and values to be able to do so. To make this case, we draw on the human capabilities approach of Sen (1999, 2009 and Nussbaum (2000, 2011 to conceptualise professional education for the public good as an ally of the struggles of people living in poverty and experiencing inequalities, expanding the well-being of people to be and to do in ways they have reason to value – to be mobile, cared for, respected, and so on. In particular we are interested in which human capabilities and functionings are most needed for a professional practice and professionalism that can contribute to transformative social change and how professional development is enabled via pedagogical arrangements.

  8. Teachers' incorporation of nanoscale science and engineering lessons into the classroom and factors that influence this incorporation

    Science.gov (United States)

    Hutchinson, Kelly M.

    . Teachers also had to believe that their students were capable of learning NSE concepts and that the concepts were relevant to their science curriculum if they were to incorporate NSE into their classrooms.

  9. IAC-1.5 - INTEGRATED ANALYSIS CAPABILITY

    Science.gov (United States)

    Vos, R. G.

    1994-01-01

    The objective of the Integrated Analysis Capability (IAC) system is to provide a highly effective, interactive analysis tool for the integrated design of large structures. IAC was developed to interface programs from the fields of structures, thermodynamics, controls, and system dynamics with an executive system and a database to yield a highly efficient multi-disciplinary system. Special attention is given to user requirements such as data handling and on-line assistance with operational features, and the ability to add new modules of the user's choice at a future date. IAC contains an executive system, a database, general utilities, interfaces to various engineering programs, and a framework for building interfaces to other programs. IAC has shown itself to be effective in automating data transfer among analysis programs. The IAC system architecture is modular in design. 1) The executive module contains an input command processor, an extensive data management system, and driver code to execute the application modules. 2) Technical modules provide standalone computational capability as well as support for various solution paths or coupled analyses. 3) Graphics and model generation modules are supplied for building and viewing models. 4) Interface modules provide for the required data flow between IAC and other modules. 5) User modules can be arbitrary executable programs or JCL procedures with no pre-defined relationship to IAC. 6) Special purpose modules are included, such as MIMIC (Model Integration via Mesh Interpolation Coefficients), which transforms field values from one model to another; LINK, which simplifies incorporation of user specific modules into IAC modules; and DATAPAC, the National Bureau of Standards statistical analysis package. The IAC database contains structured files which provide a common basis for communication between modules and the executive system, and can contain unstructured files such as NASTRAN checkpoint files, DISCOS plot files

  10. Venkatapuram's Capability theory of Health: A Critical Discussion.

    Science.gov (United States)

    Tengland, Per-Anders

    2016-01-01

    The discussion about theories of health has recently had an important new input through the work of Sridhar Venkatapuram. He proposes a combination of Lennart Nordenfelt's holistic theory of health and Martha Nussbaum's version of the capability approach. The aim of the present article is to discuss and evaluate this proposal. The article starts with a discussion of Nordenfelt's theory and evaluates Venkatapuram' critique of it, that is, of its relativism, both regarding goals and environment, and of the subjectivist theory of happiness used. Then the article explains why Nordenfelt's idea of a reasonable environment is not a problem for the theory, and it critiques Venkatapuram's own incorporation of the environment into the concept of health, suggesting that this makes the concept too wide. It contends, moreover, that Venkatapuram's alternative theory retains a problem inherent in Nordenfelt's theory, namely, that health is conceived of as a second-order ability. It is argued that health should, instead, be defined as first-order abilities. This means that health cannot be seen as a capability, and also that health cannot be seen as a meta-capability of the kind envisioned by Venkatapuram. It is, furthermore, argued that the theory lacks one crucial aspect of health, namely, subjective wellbeing. Finally, the article tries to illustrate how health, in the suggested alternative sense, as first-order abilities, fits into Nussbaum's capability theory, since health as an 'actuality' is part of all the 'combined capabilities' suggested by Nussbaum. PMID:26686326

  11. Venkatapuram's Capability theory of Health: A Critical Discussion.

    Science.gov (United States)

    Tengland, Per-Anders

    2016-01-01

    The discussion about theories of health has recently had an important new input through the work of Sridhar Venkatapuram. He proposes a combination of Lennart Nordenfelt's holistic theory of health and Martha Nussbaum's version of the capability approach. The aim of the present article is to discuss and evaluate this proposal. The article starts with a discussion of Nordenfelt's theory and evaluates Venkatapuram' critique of it, that is, of its relativism, both regarding goals and environment, and of the subjectivist theory of happiness used. Then the article explains why Nordenfelt's idea of a reasonable environment is not a problem for the theory, and it critiques Venkatapuram's own incorporation of the environment into the concept of health, suggesting that this makes the concept too wide. It contends, moreover, that Venkatapuram's alternative theory retains a problem inherent in Nordenfelt's theory, namely, that health is conceived of as a second-order ability. It is argued that health should, instead, be defined as first-order abilities. This means that health cannot be seen as a capability, and also that health cannot be seen as a meta-capability of the kind envisioned by Venkatapuram. It is, furthermore, argued that the theory lacks one crucial aspect of health, namely, subjective wellbeing. Finally, the article tries to illustrate how health, in the suggested alternative sense, as first-order abilities, fits into Nussbaum's capability theory, since health as an 'actuality' is part of all the 'combined capabilities' suggested by Nussbaum.

  12. Market dispatch incorporating stability constraints

    International Nuclear Information System (INIS)

    Stability aspects have often been incorporated in the electricity market dispatch/pricing procedure using trial-and-error methods, or approximated in the dispatch optimisation directly as a set of linear constraints on generation/transmission. This paper presents the preliminary experiences with the development of a market optimal power flow (OPF) model that incorporates both transient and voltage stability constraints. The resultant dispatch and prices are expected to exhibit the impact of accurately modelled stability limits that are hitherto largely unknown. This model allows integrated representation of both voltage and transient stability. It, however, entails very significant computational complexities. A complete resolution of all these issues is beyond the scope of this paper, although some initial thoughts to simplify computation are discussed. The importance of stability constraints on market dispatch and prices is discussed around a simple 9-bus system example. (author)

  13. JEM/SMILES observation capability

    Science.gov (United States)

    Kasai, Yasuko J.; Baron, Philippe; Ochiai, Satoshi; Mendrok, Jana; Urban, Joachim; Murtagh, Donal; Moller, Joakim; Manabe, Takeshi; Kikuchi, Kenichi; Nishibori, Toshiyuki

    2009-09-01

    A new generation of sub-millimeter-wave receivers employing sensitive SIS (Superconductor-Insulator- Superconductor) detector technology will provide new opportunities for precise passive remote sensing observation of minor constituents in atmosphere. Superconducting Submillimeter-Wave Limb-Emission Sounder (SMILES) was designed to be onbord the Japanese Experiment Module (JEM) on the International Space Station (ISS) as a collaboration project of National Institute of Information and Communications Technology (NICT) and Japan Aerospace Exploration Agency (JAXA). SMILES scheduled to be launch in September 11, 2009 by the H-II Transfer Vehicle (HTV). Mission Objectives are: i) Space demonstration of superconductive mixer and 4-K mechanical cooler for the submillimeter limb emission sounding, and ii) global observations of atmospheric minor constituents. JEM/SMILES will allow to observe the atmospheric species such as O3, H35Cl, H37 Cl, ClO, BrO, HOCl, HO2, and HNO3, CH3CN, and Ozone isotope species with the precisions in a few to several tens percents from upper troposphere to the mesosphere. We have estimated the observation capabilities of JEM/SMILES. This new technology may allow us to open new issues in atmospheric science.

  14. Solar mechanics thermal response capabilities.

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, Dean D.

    2009-07-01

    In many applications, the thermal response of structures exposed to solar heat loads is of interest. Solar mechanics governing equations were developed and integrated with the Calore thermal response code via user subroutines to provide this computational simulation capability. Solar heat loads are estimated based on the latitude and day of the year. Vector algebra is used to determine the solar loading on each face of a finite element model based on its orientation relative to the sun as the earth rotates. Atmospheric attenuation is accounted for as the optical path length varies from sunrise to sunset. Both direct and diffuse components of solar flux are calculated. In addition, shadowing of structures by other structures can be accounted for. User subroutines were also developed to provide convective and radiative boundary conditions for the diurnal variations in air temperature and effective sky temperature. These temperature boundary conditions are based on available local weather data and depend on latitude and day of the year, consistent with the solar mechanics formulation. These user subroutines, coupled with the Calore three-dimensional thermal response code, provide a complete package for addressing complex thermal problems involving solar heating. The governing equations are documented in sufficient detail to facilitate implementation into other heat transfer codes. Suggestions for improvements to the approach are offered.

  15. OPSAID improvements and capabilities report.

    Energy Technology Data Exchange (ETDEWEB)

    Halbgewachs, Ronald D.; Chavez, Adrian R.

    2011-08-01

    Process Control System (PCS) and Industrial Control System (ICS) security is critical to our national security. But there are a number of technological, economic, and educational impediments to PCS owners implementing effective security on their systems. Sandia National Laboratories has performed the research and development of the OPSAID (Open PCS Security Architecture for Interoperable Design), a project sponsored by the US Department of Energy Office of Electricity Delivery and Energy Reliability (DOE/OE), to address this issue. OPSAID is an open-source architecture for PCS/ICS security that provides a design basis for vendors to build add-on security devices for legacy systems, while providing a path forward for the development of inherently-secure PCS elements in the future. Using standardized hardware, a proof-of-concept prototype system was also developed. This report describes the improvements and capabilities that have been added to OPSAID since an initial report was released. Testing and validation of this architecture has been conducted in another project, Lemnos Interoperable Security Project, sponsored by DOE/OE and managed by the National Energy Technology Laboratory (NETL).

  16. Requirements Development for Interoperability Simulation Capability for Law Enforcement

    Energy Technology Data Exchange (ETDEWEB)

    Holter, Gregory M.

    2004-05-19

    The National Counterdrug Center (NCC) was initially authorized by Congress in FY 1999 appropriations to create a simulation-based counterdrug interoperability training capability. As the lead organization for Research and Analysis to support the NCC, the Pacific Northwest National Laboratory (PNNL) was responsible for developing the requirements for this interoperability simulation capability. These requirements were structured to address the hardware and software components of the system, as well as the deployment and use of the system. The original set of requirements was developed through a process of conducting a user-based survey of requirements for the simulation capability, coupled with an analysis of similar development efforts. The user-based approach ensured that existing concerns with respect to interoperability within the law enforcement community would be addressed. Law enforcement agencies within the designated pilot area of Cochise County, Arizona, were surveyed using interviews and ride-alongs during actual operations. The results of this survey were then accumulated, organized, and validated with the agencies to ensure the accuracy of the results. These requirements were then supplemented by adapting operational requirements from existing systems to ensure system reliability and operability. The NCC adopted a development approach providing incremental capability through the fielding of a phased series of progressively more capable versions of the system. This allowed for feedback from system users to be incorporated into subsequent revisions of the system requirements, and also allowed the addition of new elements as needed to adapt the system to broader geographic and geopolitical areas, including areas along the southwest and northwest U.S. borders. This paper addresses the processes used to develop and refine requirements for the NCC interoperability simulation capability, as well as the response of the law enforcement community to the use of

  17. A Roadmap for NEAMS Capability Transfer

    Energy Technology Data Exchange (ETDEWEB)

    Bernholdt, David E [ORNL

    2011-11-01

    The vision of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is to bring truly predictive modeling and simulation (M&S) capabilities to the nuclear engineering community in order to enable a new approach to the design and analysis of nuclear energy systems. From its inception, the NEAMS program has always envisioned a broad user base for its software and scientific products, including researchers within the DOE complex, nuclear industry technology developers and vendors, and operators. However activities to date have focused almost exclusively on interactions with NEAMS sponsors, who are also near-term users of NEAMS technologies. The task of the NEAMS Capability Transfer (CT) program element for FY2011 is to develop a comprehensive plan to support the program's needs for user outreach and technology transfer. In order to obtain community input to this plan, a 'NEAMS Capability Transfer Roadmapping Workshop' was held 4-5 April 2011 in Chattanooga, TN, and is summarized in this report. The 30 workshop participants represented the NEAMS program, the DOE and industrial user communities, and several outside programs. The workshop included a series of presentations providing an overview of the NEAMS program and presentations on the user outreach and technology transfer experiences of (1) The Advanced Simulation and Computing (ASC) program, (2) The Standardized Computer Analysis for Licensing Evaluation (SCALE) project, and (3) The Consortium for Advanced Simulation of Light Water Reactors (CASL), followed by discussion sessions. Based on the workshop and other discussions throughout the year, we make a number of recommendations of key areas for the NEAMS program to develop the user outreach and technology transfer activities: (1) Engage not only DOE, but also industrial users sooner and more often; (2) Engage with the Nuclear Regulatory Commission to facilitate their understanding and acceptance of NEAMS approach to predictive M&S; (3

  18. Incorporating Spirituality in Primary Care.

    Science.gov (United States)

    Isaac, Kathleen S; Hay, Jennifer L; Lubetkin, Erica I

    2016-06-01

    Addressing cultural competency in health care involves recognizing the diverse characteristics of the patient population and understanding how they impact patient care. Spirituality is an aspect of cultural identity that has become increasingly recognized for its potential to impact health behaviors and healthcare decision-making. We consider the complex relationship between spirituality and health, exploring the role of spirituality in primary care, and consider the inclusion of spirituality in existing models of health promotion. We discuss the feasibility of incorporating spirituality into clinical practice, offering suggestions for physicians. PMID:26832335

  19. Organizational Economics of Capability and Heterogeneity

    DEFF Research Database (Denmark)

    Argyres, Nicholas S.; Felin, Teppo; Foss, Nicolai Juul;

    2012-01-01

    For decades, the literatures on firm capabilities and organizational economics have been at odds with each other, specifically relative to explaining organizational boundaries and heterogeneity. We briefly trace the history of the relationship between the capabilities literature and organizationa...

  20. Building capabilities to manage strategic alliances

    OpenAIRE

    SLUYTS, Kim; Matthyssens, Paul; Martens, Rudy; Streukens, Sandra

    2011-01-01

    Recently, academics have attributed a large part of alliance success to a firm's ability to successfully manage its alliances, also called its level of alliance management capability. We contribute to this growing body of literature by (1) verifying the impact of alliance management capability on alliance performance and (2) analyzing the drivers of alliance management capability. We measure this capability through four types of alliance learning processes and study how each of these processe...

  1. Work, Subjective Well-being and Capabilities

    OpenAIRE

    Suppa, Nicolai

    2014-01-01

    Chapter 2: This chapter explores the link between poverty as capability deprivation and current life satisfaction. Using German panel data, I examine both whether capability deprivation does hurt and whether individuals eventually adapt. To detect capability deprivation I draw on the notion of an inadequate income together with nonconsumption data of specific commodities. Assumptions and conditions rendering this approach valid are scrutinised. The results indicate that capability...

  2. A comprehensive estimation method for enterprise capability

    OpenAIRE

    Tetiana Kuzhda; Natalia Kyrych

    2015-01-01

    In today’s highly competitive business world, the need for efficient enterprise capability management is greater than ever. As more enterprises begin to compete on a global scale, the effective use of enterprise capability will become imperative for them to improve their business activities. The definition of socio-economic capability of the enterprise has been given and the main components of enterprise capability have been pointed out. The comprehensive method to estimate enterprise capabil...

  3. Sustainability economics, ontology and the capability approach

    OpenAIRE

    Martins, Nuno O.

    2011-01-01

    Copyright © 2011 Elsevier B.V. All rights reserved. The relationship between sustainability economics and the capability approach has recently been explored. Here I shall discuss this relationship, and argue that a study of the ontology underlying the capability approach can help us to see more clearly the interconnections between sustainability economics and the capability approach. In particular, the interpretations of the capability approach as an ontological exercise, which have recent...

  4. DESIGNING AND MEASURING CAPABILITY. A NEW PERSPECTIVE

    Directory of Open Access Journals (Sweden)

    Ovidiu COCENESCU

    2013-01-01

    Full Text Available The concept of capability has long been a topic for debate among planners. The main benefit brought by it is that of ensuring the connection between objectives and necessary financial resources. Thus, capability is a median element within the process of integrated planning. In this context, there are sceptics who consider that capability cannot be measured. However, this article aims at presenting a pattern and a formula for measuring and interpreting the level of capability.

  5. Capability development within the multinational corporation

    OpenAIRE

    Kilpinen, Paula

    2013-01-01

    The operating environment is being shaped by globalization forces, rapid technological change and intensified competition, which call for strategic changes and new capabilities from multinational corporations. Even though the capability-based determinants to firm survival and growth have been recognized, research on capability development has been limited. This study investigates capability dynamics within MNCs and the interactions between strategy and the environments internal and external t...

  6. Dynamic capabilities in small software firms

    OpenAIRE

    Kivelä, Marianne

    2007-01-01

    In this dissertation we study dynamic capabilities in small software firms. Small software firms find themselves in highly complex and turbulent environments that require dynamic capabilities to build, integrate and configure resources. While the literature describes a portfolio of such dynamic capabilities that can help firms to adapt to changing conditions, we could not find many definitions, models and studies on these capabilities suitable with particular focus on small software firms. Fu...

  7. Selecting Capabilities for Quality of Life Measurement

    Science.gov (United States)

    Robeyns, Ingrid

    2005-01-01

    The capability approach advocates that interpersonal comparisons be made in the space of functionings and capabilities. However, Amartya Sen has not specified which capabilities should be selected as the relevant ones. This has provoked two types of criticism. The stronger critique is Martha Nussbaum's claim that Sen should endorse one specific…

  8. Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Lee; Park, Chan Eok; Lee, Sang Yong [Korea Atomic Energy Research Institute, Yusung, Taejon (Korea, Republic of)] [and others

    1996-10-01

    Assessment of the original REAP/N4OD3.1 code against the FLECHT SEASET series of experiments has identified some weaknesses of the reflood model, such as the lack of a quenching temperature model, the shortcoming of the Chen transition boiling model, and the incorrect prediction of droplet size and interfacial heat transfer. Also, high temperature spikes during the reflood calculation resulted in high steam flow oscillation and liquid carryover. An effort had been made to improve the code with respect to the above weakness, and the necessary model for the wall heat transfer package and the numerical scheme had been modified. Some important FLECHT-SEASET experiments were assessed using the improved version and standard version. The result from the improved REAP/MOD3.1 shows the weaknesses of REAP/N4OD3.1 were much improved when compared to the standard MOD3.1 code. The prediction of void profile and cladding temperature agreed better with test data, especially for the gravity feed test. The scatter diagram of peak cladding temperatures (PCTs) is made from the comparison of all the calculated PCTs and the corresponding experimental values. The deviation between experimental and calculated PCTs were calculated for 2793 data points. The deviations are shown to be normally distributed, and used to quantify statistically the PCT uncertainty of the code. The upper limit of PCT uncertainty at 95% confidence level is evaluated to be about 99K.

  9. Thermal-hydraulic system study of a high pressure, high temperature helium loop using RELAP5-3D code

    International Nuclear Information System (INIS)

    Highlights: ► A thermal-hydraulic system analysis for a high pressure, high temperature helium loop has been investigated. ► The loop belongs to the Helium Loop Karlsruhe (HELOKA) facility, which contains the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) as the test module. ► The loop including all components has been modeled using the system code REALP5-3D, and the main control strategy has been implemented as well. ► With this model, the loop dynamics in conditions relevant for blanket module operation have been demonstrated. - Abstract: The thermal-hydraulic system analysis for the Helium Loop Karlsruhe (HELOKA) facility, a high pressure, high temperature experimental helium loop having the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) as the test module, was investigated. Using the system code REALP5-3D, all components in the loop are modeled as well as the main control strategy. With this model, the loop dynamics in conditions relevant for blanket module operation are simulated and analyzed.

  10. Summary of severe accident assessment for Atucha 2 Nuclear Power Plant using RELAP5/SCDAPSIM Mod3.6

    International Nuclear Information System (INIS)

    A severe accident assessment was performed for the Atucha 2 Nuclear Power Plant in Argentina. Atucha 2 is a PHWR, cooled and moderated by heavy water, presently in commissioning process. Its 451 fuel assemblies are 6.03m high and each composed of 37 Zircaloy clad fuel rods. Each assembly is placed inside an individual Zircaloy coolant channel. Heavy water coolant flows inside the channels which are all immersed inside the moderator tank. The RPV lower plenum is occupied by a massive steel structure called 'filling body' that was designed to minimize heavy water inventory. Due to some unique design characteristics, severe accident progression in Atucha 2 is expected to be somewhat different from that predicted for regular PWRs. Therefore, a very detailed assessment was performed, focused on the different accident stages and expected phenomena by the use of different input models and nodalizations. When possible, linking to available experimental data was performed. RELAP/SCDAPSIM Mod 3.6 was the computer code selected to perform this task. The modeling of Atucha 2's unique characteristics required several extensions to the code. For the severe accident assessment of Atucha 2, three different input models were developed that were key instruments for the debugging and evaluation process. A Single Channel Model was used to evaluate the first stages of core heatup (including the boiloff of the channels and moderator tank), an RPV standalone model was used to assess the interaction between components in the complete core and for the evaluation of late in-core melting and relocation. Then, a Lower Plenum standalone model was developed to assess the behavior of the melted and slumped core material on top of the filling body and to analyze ex-vessel cooling as a possible severe accident management action. For each of the cases, highlights of key results are shown and general conclusions are drawn. In the case of a severe accident with significant meltdown of the reactor core, the melted fuel slumps on top of the filler steel in the lower plenum. The slumped fuel has the configuration of a relatively thin plate, which affects its cooling in a beneficial manner. A flooded containment was calculated to keep the reactor vessel cool enough to maintain its structural integrity. This information will be used in the future for the full-plant model that will be developed to perform severe accident management studies. (author)

  11. Margin for In-Vessel Retention in the APR1400 - VESTA and SCDAP/RELAP5-3D Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Joy Rempe; D. Knudson

    2004-12-01

    If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with such plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe pressurized water reactor (PWR) (AP600), which relied upon external reactor vessel cooling (ERVC) for in-vessel retention (IVR), resulted in the U.S. Nuclear Regulatory Commission (USNRC) approving the design without requiring certain conventional features common to existing light water reactors (LWRs). IVR of core melt is therefore a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced LWRs. However, it is not clear that currently proposed ERVC without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a three-year, United States (U.S.) -Korean International Nuclear Energy Research Initiative (INERI) project was initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) explored options, such as enhanced ERVC performance and an enhanced in-vessel core catcher (IVCC), that have the potential to ensure that IVR is feasible for higher power reactors.

  12. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    OpenAIRE

    Andrej Prošek; Leon Cizelj

    2013-01-01

    Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pump...

  13. A comprehensive estimation method for enterprise capability

    Directory of Open Access Journals (Sweden)

    Tetiana Kuzhda

    2015-11-01

    Full Text Available In today’s highly competitive business world, the need for efficient enterprise capability management is greater than ever. As more enterprises begin to compete on a global scale, the effective use of enterprise capability will become imperative for them to improve their business activities. The definition of socio-economic capability of the enterprise has been given and the main components of enterprise capability have been pointed out. The comprehensive method to estimate enterprise capability that takes into account both social and economic components has been offered. The methodical approach concerning integrated estimation of the enterprise capability has been developed. Novelty deals with the inclusion of summary measure of the social component of enterprise capability to define the integrated index of enterprise capability. The practical significance of methodological approach is that the method allows assessing the enterprise capability comprehensively through combining two kinds of estimates – social and economic and converts them into a single integrated indicator. It provides a comprehensive approach to socio-economic estimation of enterprise capability, sets a formal basis for making decisions and helps allocate enterprise resources reasonably. Practical implementation of this method will affect the current condition and trends of the enterprise, help to make forecasts and plans for its development and capability efficient use.

  14. Distinctive Dynamic Capabilities for New Business Creation

    DEFF Research Database (Denmark)

    Rosenø, Axel; Enkel, Ellen; Mezger, Florian

    2013-01-01

    and fast-paced industries, and that similarities exist across industries. Hence, the study contributes to dynamic capabilities literature by: 1) identifying the distinctive dynamic capabilities for new business creation; 2) shifting focus away from dynamic capabilities in environments characterised by high...... clock-speed and uncertainty towards considering dynamic capabilities for the purpose of developing new businesses, which also implies a high degree of uncertainty. Based on interviews with 33 companies, we identify distinctive dynamic capabilities for new business creation, find that dynamic......This study examines the distinctive dynamic capabilities for new business creation in established companies. We argue that these are very different from those for managing incremental innovation within a company's core business. We also propose that such capabilities are needed in both slow...

  15. Activity-based resource capability modeling

    Institute of Scientific and Technical Information of China (English)

    CHENG Shao-wu; XU Xiao-fei; WANG Gang; SUN Xue-dong

    2008-01-01

    To analyse and optimize a enterprise process in a wide scope, an activity-based method of modeling resource capabilities is presented. It models resource capabilities by means of the same structure as an activity, that is, resource capabilities are defined by input objects, actions and output objects. A set of activity-based re-source capability modeling rules and matching rules between an activity and a resource are introduced. This method can not only be used to describe capability of manufacturing tools, but also capability of persons and applications, etc. It unifies methods of modeling capability of all kinds of resources in an enterprise and supports the optimization of the resource allocation of a process.

  16. Predicting Mortgage Arrears: An Investigation Into the Predictive Capability of Customer Spending Patterns

    OpenAIRE

    Roche, Jamie

    2014-01-01

    The management of credit risk and mortgage arrears has become a very important area in financial services and banking. This dissertation set out to build a statistical model, which incorporates customer spending habits and the current equity value of a property, capable of predicting arrears. Current literature identifies many themes such as negative equity and unemployment that are common occurring factors in mortgage arrears but a multi-faceted approach was required to build a model capable...

  17. Numeral Incorporation in Japanese Sign Language

    Science.gov (United States)

    Ktejik, Mish

    2013-01-01

    This article explores the morphological process of numeral incorporation in Japanese Sign Language. Numeral incorporation is defined and the available research on numeral incorporation in signed language is discussed. The numeral signs in Japanese Sign Language are then introduced and followed by an explanation of the numeral morphemes which are…

  18. 49 CFR 572.190 - Incorporated materials.

    Science.gov (United States)

    2010-10-01

    ... Register approved the materials incorporated by reference in accordance with 5 U.S.C. 552(a) and 1 CFR part... 49 Transportation 7 2010-10-01 2010-10-01 false Incorporated materials. 572.190 Section 572.190... Dummy, Small Adult Female § 572.190 Incorporated materials. (a) The following materials are...

  19. 49 CFR 572.30 - Incorporated materials.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Incorporated materials. 572.30 Section 572.30....30 Incorporated materials. (a) The drawings and specifications referred to in this regulation that... Federal Register has approved the materials incorporated by reference. For materials subject to...

  20. 49 CFR 587.5 - Incorporated materials.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Incorporated materials. 587.5 Section 587.5... Barrier § 587.5 Incorporated materials. (a) The drawings and specifications referred to in this regulation that are not set forth in full are hereby incorporated in this part by reference. These materials...

  1. Developing Flexible, High Performance Polymers with Self-Healing Capabilities

    Science.gov (United States)

    Jolley, Scott T.; Williams, Martha K.; Gibson, Tracy L.; Caraccio, Anne J.

    2011-01-01

    Flexible, high performance polymers such as polyimides are often employed in aerospace applications. They typically find uses in areas where improved physical characteristics such as fire resistance, long term thermal stability, and solvent resistance are required. It is anticipated that such polymers could find uses in future long duration exploration missions as well. Their use would be even more advantageous if self-healing capability or mechanisms could be incorporated into these polymers. Such innovative approaches are currently being studied at the NASA Kennedy Space Center for use in high performance wiring systems or inflatable and habitation structures. Self-healing or self-sealing capability would significantly reduce maintenance requirements, and increase the safety and reliability performance of the systems into which these polymers would be incorporated. Many unique challenges need to be overcome in order to incorporate a self-healing mechanism into flexible, high performance polymers. Significant research into the incorporation of a self-healing mechanism into structural composites has been carried out over the past decade by a number of groups, notable among them being the University of I1linois [I]. Various mechanisms for the introduction of self-healing have been investigated. Examples of these are: 1) Microcapsule-based healant delivery. 2) Vascular network delivery. 3) Damage induced triggering of latent substrate properties. Successful self-healing has been demonstrated in structural epoxy systems with almost complete reestablishment of composite strength being achieved through the use of microcapsulation technology. However, the incorporation of a self-healing mechanism into a system in which the material is flexible, or a thin film, is much more challenging. In the case of using microencapsulation, healant core content must be small enough to reside in films less than 0.1 millimeters thick, and must overcome significant capillary and surface

  2. Development of Approaches for Deuterium Incorporation in Plants

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Barbara R [ORNL

    2015-01-01

    Soon after the discovery of deuterium, efforts to utilize this stable isotope of hydrogen for labeling of plants began and have proven successful for natural abundance to 20% enrichment. However, isotopic labeling with deuterium (2H) in higher plants at the level of 40% and higher is complicated by both physiological responses, particularly water exchange through transpiration, and inhibitory effects of D2O on germination, rooting, and growth. The highest incorporation of 40 50% had been reported for photoheterotrophic cultivation of the duckweed Lemna. Higher substitution is desirable for certain applications using neutron scattering and nuclear magnetic resonance (NMR) techniques. 1H2H-NMR and mass spectroscopy are standard methods frequently used for determination of location and amount of deuterium substitution. The changes in infrared (IR) absorption observed for H to D substitution in hydroxyl and alkyl groups provide rapid initial evaluation of incorporation. Short-term experiments with cold-tolerant annual grasses can be carried out in enclosed growth containers to evaluate incorporation. Growth in individual chambers under continuous air perfusion with dried sterile-filtered air enables long-term cultivation of multiple plants at different D2O concentrations. Vegetative propagation from cuttings extends capabilities to species with low germination rates. Cultivation in 50% D2O of annual ryegrass and switchgrass following establishment of roots by growth in H2O produces samples with normal morphology and 30 40 % deuterium incorporation in the biomass. Winter grain rye (Secale cereale) was found to efficiently incorporate deuterium by photosynthetic fixation from 50% D2O but did not incorporate deuterated phenylalanine-d8 from the growth medium.

  3. Compilation of Sandia Laboratories technical capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Lundergan, C. D.; Mead, P. L. [eds.

    1975-11-01

    This report is a compilation of 17 individual documents that together summarize the technical capabilities of Sandia Laboratories. Each document in this compilation contains details about a specific area of capability. Examples of application of the capability to research and development problems are provided. An eighteenth document summarizes the content of the other seventeen. Each of these documents was issued with a separate report number (SAND 74-0073A through SAND 74-0091, except -0078). (RWR)

  4. Compilation of Sandia Laboratories technical capabilities

    International Nuclear Information System (INIS)

    This report is a compilation of 17 individual documents that together summarize the technical capabilities of Sandia Laboratories. Each document in this compilation contains details about a specific area of capability. Examples of application of the capability to research and development problems are provided. An eighteenth document summarizes the content of the other seventeen. Each of these documents was issued with a separate report number (SAND 74-0073A through SAND 74-0091, except -0078)

  5. Understanding dynamic capabilities through knowledge management

    DEFF Research Database (Denmark)

    Nielsen, Anders Paarup

    2006-01-01

    In the paper eight knowledge management activities are identified; knowledge creation, acquisition, capture, assembly, sharing, integration, leverage and exploitation. These activities are assembled into the three dynamic capabilities of knowledge development, knowledge (re......)combination and knowledge use. The dynamic capabilities and the associated knowledge management activities create flows to and from the firm’s stock of knowledge and they support the creation and use of organizational capabilities....

  6. Capabilities, social rights and European market integration

    OpenAIRE

    Jude Browne; Simon Deakin; Frank Wilkinson

    2002-01-01

    This paper explores the links between the economic notion of 'capabilities' and the judicial concept of social rights. We begin by revisiting TH Marshall's classic analysis of social rights and their ambiguous relationship to the market. We then examine how far Amartya Sen's Capabilities Approach provides a framework for locating social rights within a market setting. We argue that Sen's non-dogmatic, context-orientated approach to defining the meaning of capabilities offers a viable way forw...

  7. COMPLEMENTARITY OF INFORMATION TECHNOLOGY AND MARKETING CAPABILITIES

    OpenAIRE

    RAUL ALEXANDRU HUȚU

    2015-01-01

    This paper study the relation between information technology and marketing capabilities. The interface between marketing and information technology forms the subject of numerous empirical and conceptual research. The expanding adoption of information technology in marketing implies the knowledge about the e-marketing capabilities development factors and the potential of these capabilities to generate competitive advantage and to improve firm’s performance. The results of recent st...

  8. An Institutional View of Organizational Capabilities

    OpenAIRE

    Reinert do Nascimento, Mauricio; Bandeira-De-Mello, Rodrigo

    2008-01-01

    The literature of organizational capabilities has overlooked the role of values. We present an institutionalist view of the capability phenomenon accounting for the values occurring in three levels: in the action at routine level, at the organization level and at the institutional environment level. The mutual influences among these three levels are needed to analyze capability dynamics in its entirety. Central to our argument is the Weberian notion of social action and rationality, and the c...

  9. Ensemble learning incorporating uncertain registration.

    Science.gov (United States)

    Simpson, Ivor J A; Woolrich, Mark W; Andersson, Jesper L R; Groves, Adrian R; Schnabel, Julia A

    2013-04-01

    This paper proposes a novel approach for improving the accuracy of statistical prediction methods in spatially normalized analysis. This is achieved by incorporating registration uncertainty into an ensemble learning scheme. A probabilistic registration method is used to estimate a distribution of probable mappings between subject and atlas space. This allows the estimation of the distribution of spatially normalized feature data, e.g., grey matter probability maps. From this distribution, samples are drawn for use as training examples. This allows the creation of multiple predictors, which are subsequently combined using an ensemble learning approach. Furthermore, extra testing samples can be generated to measure the uncertainty of prediction. This is applied to separating subjects with Alzheimer's disease from normal controls using a linear support vector machine on a region of interest in magnetic resonance images of the brain. We show that our proposed method leads to an improvement in discrimination using voxel-based morphometry and deformation tensor-based morphometry over bootstrap aggregating, a common ensemble learning framework. The proposed approach also generates more reasonable soft-classification predictions than bootstrap aggregating. We expect that this approach could be applied to other statistical prediction tasks where registration is important. PMID:23288332

  10. Aerodynamics Laboratory Facilities, Equipment, and Capabilities

    Data.gov (United States)

    Federal Laboratory Consortium — The following facilities, equipment, and capabilities are available in the Aerodynamics Laboratory Facilities and Equipment (1) Subsonic, open-jet wind tunnel with...

  11. Esourcing capability model for service providers

    CERN Document Server

    Hefley, Keith M Heston and Bill; Hyder, Elaine

    2010-01-01

    The eSourcing Capability Model for Service Providers (eSCM-SP) is the best practices model that supports sourcing organizations successfully manage and reduce their risks and improve their capabilities across the entire sourcing life-cycle. It addresses the critical issues related to IT-enabled sourcing (eSourcing) for both outsourced and in-sourced (shared services) agreements. Each of the Model's 84 Practice is distributed along three easy to follow dimensions: Sourcing Life-cycle, Capability Area, and Capability Level, and have been applied in IT, BPO, and KPO settings.The eSCM-SP has been

  12. Methods of ecological capability evaluation of forest

    International Nuclear Information System (INIS)

    In this research common methods of ecological capability evaluation of forests were reviewed and limitations for performance were analysed. Ecological capability of forests is an index that show site potential in several role of wood production, soil conservation, flood control, biodiversity, conservation and water supply. This index is related to ecological characteristics of land, such as soil, micro climate, elevation, slope and aspect that affect potential of sites. Suitable method of ecological capability evaluation must be chosen according to the objective of forestry. Common methods for ecological capability evaluation include plant and animal diversity, site index curve, soil and land form, inter branches, index plants, leave analyses, analyses regeneration and ecological mapping

  13. Organizational Capabilities of the Entrepreneurial University

    Directory of Open Access Journals (Sweden)

    Lucian Gramescu

    2015-05-01

    Full Text Available Developing entrepreneurial capabilities has become a key competitiveness strategy in business across the world. Overall, organizational capabilities can provide performance improvements by taking an integrated approach to people, infrastructure and processes as means of codifying organizational learning. The paper proposes “organizational capability” as a valuable tool for universities who seek to develop their competitiveness entrepreneurially, especially across the EU, where higher education is no longer a guarantee for employment and alternatives are sorely needed. For this purpose, we explore conceptualizations of organizational capabilities, propose an integrative model and apply it to learn more about the development of capability from practice at Aalto University in Finland.

  14. Mission Adaptive Uas Capabilities for Earth Science and Resource Assessment

    Science.gov (United States)

    Dunagan, S.; Fladeland, M.; Ippolito, C.; Knudson, M.; Young, Z.

    2015-04-01

    Unmanned aircraft systems (UAS) are important assets for accessing high risk airspace and incorporate technologies for sensor coordination, onboard processing, tele-communication, unconventional flight control, and ground based monitoring and optimization. These capabilities permit adaptive mission management in the face of complex requirements and chaotic external influences. NASA Ames Research Center has led a number of Earth science remote sensing missions directed at the assessment of natural resources and here we describe two resource mapping problems having mission characteristics requiring a mission adaptive capability extensible to other resource assessment challenges. One example involves the requirement for careful control over solar angle geometry for passive reflectance measurements. This constraint exists when collecting imaging spectroscopy data over vegetation for time series analysis or for the coastal ocean where solar angle combines with sea state to produce surface glint that can obscure the signal. Furthermore, the primary flight control imperative to minimize tracking error should compromise with the requirement to minimize aircraft motion artifacts in the spatial measurement distribution. A second example involves mapping of natural resources in the Earth's crust using precision magnetometry. In this case the vehicle flight path must be oriented to optimize magnetic flux gradients over a spatial domain having continually emerging features, while optimizing the efficiency of the spatial mapping task. These requirements were highlighted in recent Earth Science missions including the OCEANIA mission directed at improving the capability for spectral and radiometric reflectance measurements in the coastal ocean, and the Surprise Valley Mission directed at mapping sub-surface mineral composition and faults, using high-sensitivity magnetometry. This paper reports the development of specific aircraft control approaches to incorporate the unusual and

  15. Do Acquirer Capabilities Affect Acquisition Performance? Examining Strategic and Effectiveness Capabilities in Acquirers

    OpenAIRE

    Mudde, Paul A.; Brush, Thomas

    2006-01-01

    This paper examines acquisition performance from the perspective of acquirer capabilities. It argues that the strategic capabilities underpinning a firm’s competitive strategy can be utilized to create economic value in acquisitions. Acquirers with strong cost leadership capabilities are expected to leverage these capabilities to reduce post-acquisition costs as they integrate acquisition targets. Acquirers with strong differentiation capabilities are expected to utilize their strategic capab...

  16. Mars Missions Using Emerging Commercial Space Transportation Capabilities

    Science.gov (United States)

    Gonzales, Andrew A.

    2016-01-01

    New Discoveries regarding the Martian Environment may impact Mars mission planning. Transportation of investigation payloads can be facilitated by Commercial Space Transportation options. The development of Commercial Space Transportation. Capabilities anticipated from various commercial entities are examined objectively. The potential for one of these options, in the form of a Mars Sample Return mission, described in the results of previous work, is presented to demonstrate a high capability potential. The transportation needs of the Mars Environment Team Project at ISU 2016 may fit within the payload capabilities of a Mars Sample Return mission, but the payload elements may or may not differ. Resource Modules will help you develop a component of a strategy to address the Implications of New Discoveries in the Martian Environment using the possibility of efficient, commercial space transportation options. Opportunities for open discussions as appropriate during the team project formulation period at the end of each Resource Module. The objective is to provide information that can be incorporated into your work in the Team Project including brainstorming.

  17. Mutagenic effect of radionuclides incorporated into DNA of Drosophila melanogaster. Progress report, 1978-1979

    International Nuclear Information System (INIS)

    Current progress in studies on the mutagenic effect of 3H incorporated into the DNA of Drosophila melanogaster is reported. It was shown that selected 3H precursors incorporated into DNA are metabolized. The forms (metabolites) of tritium found in the DNA molecules and the mutation frequencies resulting therefrom were identified. An alcohol dehydrogenase system was developed for recovering mutations that is capable of distinguishing between base changes and chain breakage events that may lead to the formation of deletions

  18. Methodological Individualism and the Organizational Capabilities Approach

    DEFF Research Database (Denmark)

    Felin, Teppo; Foss, Nicolai Juul

    2004-01-01

    critical individual-levelconsiderations, including individual action and heterogeneity. In this note we do not denyor reject the notion of routines or capabilities per se, but rather call for an increasedemphasis on how these collective structures originate and change as a result of individualactions.......Key words: Organizational capabilities, methodological individualism, philosophy ofsocial science...

  19. A framework for offshore vendor capability development

    Science.gov (United States)

    Yusuf Wibisono, Yogi; Govindaraju, Rajesri; Irianto, Dradjad; Sudirman, Iman

    2016-02-01

    Offshore outsourcing is a common practice conducted by companies, especially in developed countries, by relocating one or more their business processes to other companies abroad, especially in developing countries. This practice grows rapidly owing to the ease of accessing qualified vendors with a lower cost. Vendors in developing countries compete more intensely to acquire offshore projects. Indonesia is still below India, China, Malaysia as main global offshore destinations. Vendor capability is among other factors that contribute to the inability of Indonesian vendor in competing with other companies in the global market. Therefore, it is essential to study how to increase the vendor's capability in Indonesia, in the context of global offshore outsourcing. Previous studies on the vendor's capability mainly focus on capabilities without considering the dynamic of capabilities due to the environmental changes. In order to be able to compete with competitors and maintain the competitive advantage, it is necessary for vendors to develop their capabilities continuously. The purpose of this study is to develop a framework that describes offshore vendor capability development along the client-vendor relationship stages. The framework consists of three main components, i.e. the stages of client-vendor relationship, the success of each stage, and the capabilities of vendor at each stage.

  20. Knowledge Perspectives on Advancing Dynamic Capability

    NARCIS (Netherlands)

    van Reijsen, J.

    2014-01-01

    Dynamic Capability is the organizational capacity to timely adapt to a changing market environment by reconfiguring resources and routines in order to stay competitive. Although dynamic capability is considered the Holy Grail of strategic management, a connection to the knowledge management domain i