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Sample records for analysis code trac

  1. Development of a graphical user interface for the TRAC plant/safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, A.E.; Harkins, C.K.; Smith, R.J.

    1995-09-01

    A graphical user interface (GUI) for the Transient Reactor Analysis Code (TRAC) has been developed at Knolls Atomic Power Laboratory. This X Window based GUI supports the design and analysis process, acting as a preprocessor, runtime editor, help system and post processor to TRAC-PF1/MOD2. TRAC was developed at the Los Alamos National Laboratory (LANL). The preprocessor is an icon-based interface which allows the user to create a TRAC model. When the model is complete, the runtime editor provides the capability to execute and monitor TRAC runs on the workstation or supercomputer. After runs are made, the output processor allows the user to extract and format data from the TRAC graphics file. The TRAC GUI is currently compatible with TRAC-PF1/MOD2 V5.3 and is available with documentation from George Niederauer, Section Leader of the Software Development Section, Group TSA-8, at LANL. Users may become functional in creating, running, and interpreting results from TRAC without having to know Unix commands and the detailed format of any of the data files. This reduces model development and debug time and increases quality control. Integration with post-processing and visualization tools increases engineering effectiveness.

  2. Development of a graphical user interface for the TRAC plant/safety analysis code

    International Nuclear Information System (INIS)

    Kelly, A.E.; Harkins, C.K.; Smith, R.J.

    1995-01-01

    A graphical user interface (GUI) for the Transient Reactor Analysis Code (TRAC) has been developed at Knolls Atomic Power Laboratory. This X Window based GUI supports the design and analysis process, acting as a preprocessor, runtime editor, help system and post processor to TRAC-PF1/MOD2. TRAC was developed at the Los Alamos National Laboratory (LANL). The preprocessor is an icon-based interface which allows the user to create a TRAC model. When the model is complete, the runtime editor provides the capability to execute and monitor TRAC runs on the workstation or supercomputer. After runs are made, the output processor allows the user to extract and format data from the TRAC graphics file. The TRAC GUI is currently compatible with TRAC-PF1/MOD2 V5.3 and is available with documentation from George Niederauer, Section Leader of the Software Development Section, Group TSA-8, at LANL. Users may become functional in creating, running, and interpreting results from TRAC without having to know Unix commands and the detailed format of any of the data files. This reduces model development and debug time and increases quality control. Integration with post-processing and visualization tools increases engineering effectiveness

  3. Analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1 code

    International Nuclear Information System (INIS)

    Ajima, Toshio; Kurihara, Ryoichi; Seki, Yasushi

    1999-08-01

    The Transient Reactor Analysis Code (TRAC-BF1) was modified on the basis of ICE experimental results so as to analyze the Ingress of Coolant Event (ICE) in the vacuum vessel of a nuclear fusion reactor. In the previous report, the TRAC-BF1 code, which was originally developed for the safety analysis of a light water reactor, had been modified for the ICE of the fusion reactor. And the addition of the flat structural plate model to the VESSEL component and arbitrary appointment of the gravity direction had been added in the TRAC-BF1 code. This TRAC-BF1 code was further modified. The flat structural plate model of the VESSEL component was enabled to divide in multi layers having different materials, and a part of the multi layers could take a buried heater into consideration. Moreover, the TRAC-BF1 code was modified to analyze under the low-pressure condition close to vacuum within range of the steam table. This paper describes additional functions of the modified TRAC-BF1 code, analytical evaluation using ICE experimental data and the ITER model with final design report (FDR) data. (author)

  4. Study plan for the sensitivity analysis of the Terrain-Responsive Atmospheric Code (TRAC)

    International Nuclear Information System (INIS)

    Restrepo, L.F.; Deitesfeld, C.A.

    1987-01-01

    Rocky Flats Plant, Golden, Colorado is presently developing a computer code to model the dispersion of potential or actual releases of radioactive or toxic materials to the environment, along with the public consequences from these releases. The model, the Terrain-Responsive Atmospheric Code (TRAC), considers several complex features which could affect the overall dispersion and consequences. To help validate TRAC, a sensitivity analysis is being planned to determine how sensitive the model's solutions are to input variables. This report contains a brief description of the code, along with a list of tasks and resources needed to complete the sensitivity analysis

  5. TRAC-BD1: transient reactor analysis code for boiling-water systems

    International Nuclear Information System (INIS)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

  6. Development of Transient-Reactor Analysis Code (TRAC) for real-time applications

    International Nuclear Information System (INIS)

    Niederauer, G.F.; Giguere, P.T.; Lime, J.F.; Knight, T.D.; Ashy, O.; Fakory, R.

    1997-01-01

    This is the final report of a six-month, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Nuclear-plant training simulators employ simplified one-dimensional thermal-hydraulics codes because of the demands to run in real time and with limited computing power. The objective of this project was to investigate the feasibility of using the advanced Transient-Reactor Analysis Code (TRAC) in a simulator to increase the fidelity of a simulator. Many issues need to be addressed to take such a complex code from a batch engineering environment to a real-time environment. Working with simulator vendor, GSE, the authors investigated the technical issues relating to integrating TRAC into a real-time environment. They also modified a nuclear power plant model for simulator purposes and investigated its performance in real time

  7. Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

    International Nuclear Information System (INIS)

    Mahaffy, J.; Boyack, B.E.; Steinke, R.G.

    1998-05-01

    The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization

  8. TRAC-CFD code integration and its application to containment analysis

    International Nuclear Information System (INIS)

    Tahara, M.; Arai, K.; Oikawa, H.

    2004-01-01

    Several safety systems utilizing natural driving force have been recently adopted for operating reactors, or applied to next-generation reactor design. Examples of these safety systems are the Passive Containment Cooling System (PCCS) and the Drywell Cooler (DWC) for removing decay heat, and the Passive Auto-catalytic Recombiner (PAR) for removing flammable gas in reactor containment during an accident. DWC is used in almost all Boiling Water Reactors (BWR) in service. PAR has been introduced for some reactors in Europe and will be introduced for Japanese reactors. PCCS is a safety device of next-generation BWR. The functional mechanism of these safety systems is closely related to the transient of the thermal-hydraulic condition of the containment atmosphere. The performance depends on the containment atmospheric condition, which is eventually affected by the mass and energy changes caused by the safety system. Therefore, the thermal fluid dynamics in the containment vessel should be appropriately considered in detail to properly estimate the performance of these systems. A computational fluid dynamics (CFD) code is useful for evaluating detailed thermal hydraulic behavior related to this equipment. However, it also requires a considerable amount of computational resources when it is applied to whole containment system transient analysis. The paper describes the method and structure of the integrated analysis tool, and discusses the results of its application to the start-up behavior analysis of a containment cooling system, a drywell local cooler. The integrated analysis code was developed and applied to estimate the DWC performance during a severe accident. The integrated analysis tool is composed of three codes, TRAC-PCV, CFD-DW and TRAC-CC, and analyzes the interaction of the natural convection and steam condensation of the DWC as well as analyzing the thermal hydraulic transient behavior of the containment vessel during a severe accident in detail. The

  9. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  10. TRAC code development status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Liles, D.R.; Nelson, R.A.

    1986-01-01

    This report summarizes the characteristics and current status of the TRAC-PF1/MOD1 computer code. Recent error corrections and user-convenience features are described, and several user enhancements are identified. Current plans for the release of the TRAC-PF1/MOD2 computer code and some preliminary MOD2 results are presented. This new version of the TRAC code implements stability-enhancing two-step numerics into the 3-D vessel, using partial vectorization to obtain a code that has run 400% faster than the MOD1 code

  11. Pre-Test Analysis of the MEGAPIE Spallation Source Target Cooling Loop Using the TRAC/AAA Code

    International Nuclear Information System (INIS)

    Bubelis, Evaldas; Coddington, Paul; Leung, Waihung

    2006-01-01

    A pilot project is being undertaken at the Paul Scherrer Institute in Switzerland to test the feasibility of installing a Lead-Bismuth Eutectic (LBE) spallation target in the SINQ facility. Efforts are coordinated under the MEGAPIE project, the main objectives of which are to design, build, operate and decommission a 1 MW spallation neutron source. The technology and experience of building and operating a high power spallation target are of general interest in the design of an Accelerator Driven System (ADS) and in this context MEGAPIE is one of the key experiments. The target cooling is one of the important aspects of the target system design that needs to be studied in detail. Calculations were performed previously using the RELAP5/Mod 3.2.2 and ATHLET codes, but in order to verify the previous code results and to provide another capability to model LBE systems, a similar study of the MEGAPIE target cooling system has been conducted with the TRAC/AAA code. In this paper a comparison is presented for the steady-state results obtained using the above codes. Analysis of transients, such as unregulated cooling of the target, loss of heat sink, the main electro-magnetic pump trip of the LBE loop and unprotected proton beam trip, were studied with TRAC/AAA and compared to those obtained earlier using RELAP5/Mod 3.2.2. This work extends the existing validation data-base of TRAC/AAA to heavy liquid metal systems and comprises the first part of the TRAC/AAA code validation study for LBE systems based on data from the MEGAPIE test facility and corresponding inter-code comparisons. (authors)

  12. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    International Nuclear Information System (INIS)

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-01-01

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  15. TRAC-PF1/MOD1 computer code

    International Nuclear Information System (INIS)

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

    1984-01-01

    The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A refined version, called TRAC-P1A, was released to the National Energy Software Center (NESC) in March 1979. Although it still treats the same class of problems, TRAC-P1A is more efficient than TRAC-P1 and incorporates improved hydrodynamic and heat-transfer models. It also is easier to implement on various computers. TRAC-PD2 contains improved reflood and heat-transfer models and improvements in the numerical solution methods. Although a large LOCA code, it has been applied successfully to small-break problems and to the Three Mile Island incident. Distinguishing characteristics of the TRAC-PF1/MOD1 are summarized

  16. Numerical analysis on ingress-of-coolant events in fusion reactors with TRAC-PF1 code

    International Nuclear Information System (INIS)

    Ose, Yasuo; Takase, Kazuyuki; Akimoto, Hajime

    2000-01-01

    As for accident events related with thermal-hydraulics, in a fusion experimental reactor an ingress-of-coolant event (ICE) and a loss-of-vacuum-accident event (LOVA) should be considered. An integrated ICE/LOVA test apparatus is under planning in order to estimate quantitatively heat transfer and fluid flow characteristics under ICE and LOVA events. This study was carried out to predict numerically the thermal-hydraulic characteristics in fusion reactors at the ICE events before construction of the integrated ICE/LOVA test apparatus. The TRAC-PF1 code, which was originally developed for the thermal-hydraulic safety analysis in light water reactors, was used. The numerical analyses were performed for two kinds of system configuration with/without a pressure-suppression tank:the former for is investigation of the pressure rise characteristics and two-phase flow behavior; the latter for estimation of an effect of the pressure reduction due to the pressure-suppression tank. From the present analytical results, effects of the ingress water flow rate and vessel temperatures on the pressure rise ware clarified quantitatively. Furthermore, the pressure-rise suppression effect due to the vapor condensation in the pressure-suppression tank was predicted numerically. In addition, the useful information regarding to the design of the integrated ICE/LOVA test apparatus and the knowledge with respect to the effective usage of the TRAC-PF1 code were obtained through the present numerical study. (author)

  17. Post-processing of the TRAC code's results

    International Nuclear Information System (INIS)

    Baron, J.H.; Neuman, D.

    1987-01-01

    The TRAC code serves for the analysis of accidents in nuclear installations from the thermohydraulic point of view. A program has been developed with the aim of processing the information rapidly generated by the code, with screening graph capacity, both in high and low resolution, or either in paper through printer or plotter. Although the programs are intended to be used after the TRAC runs, they may be also used even when the program is running so as to observe the calculation process. The advantages of employing this type of tool, its actual capacity and its possibilities of expansion according to the user's needs are herein described. (Author)

  18. Development of REFLA/TRAC code for engineering work station

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Akimoto, Hajime; Murao, Yoshio

    1994-03-01

    The REFLA/TRAC code is a best-estimate code which is expected to check reactor safety analysis codes for light water reactors (LWRs) and to perform accident analyses for LWRs and also for an advanced LWR. Therefore, a high predictive capability is required and the assessment of each physical model becomes important because the models govern the predictive capability. In the case of the assessment of three-dimensional models in REFLA/TRAC code, a conventional large computer is being used and it is difficult to perform the assessment efficiently because the turnaround time for the calculation and the analysis is long. Then, a REFLA/TRAC code which can run on an engineering work station (EWS) was developed. Calculational speed of the current EWS is the same order as that of large computers and the EWS has an excellent function for multidimensional graphical drawings. Besides, the plotting processors for X-Y drawing and for two-dimensional graphical drawing were developed in order to perform efficient analyses for three-dimensional calculations. In future, we can expect that the assessment of three-dimensional models becomes more efficient by introducing an EWS with higher calculational speed and with improved graphical drawings. In this report, each outline for the following three programs is described: (1) EWS version of REFLA/TRAC code, (2) Plot processor for X-Y drawing and (3) Plot processor for two-dimensional graphical drawing. (author)

  19. Turbine trip transient analysis in peach bottom NPP with TRAC-BF1 code and Simtab-1D methodology

    International Nuclear Information System (INIS)

    Barrachina, T.; Miro, R.; Verdu, G.; Collazo, I.; Gonzalez, P.; Concejal, A.; Ortego, P.; Melara, J.

    2010-01-01

    In TRAC-BF1 nuclear cross-sections are specified in the input deck in as a polynomial expansion. Therefore, it is necessary to obtain the coefficients of this polynomial function. One of the methods proposed in the literature is the KINPAR methodology. This methodology uses the results from different perturbations of the original state to obtain the coefficients of the polynominal expansion. The simulations are performed using the SIMULATE3 code. In this work, a new methodology to obtain the cross-sections set in 1D is presented. The first step consists of the application of the SIMTAB methodology, developed in UPV, to obtain the 3D cross-sections sets from CASMO4/SIMULATE3. These 3D cross-sections sets are collapsed to 1D, using as a weighting factor the 3D thermal and rapid neutron fluxes obtained from SIMULATE3. The 1D cross-sections obtained are in the same format as the 3D sets, hence, it has been necessary to modify the TRAC-BF1 code in order to be able to read and interpolate between these tabulated 1D cross-sections. With this new methodology it is not necessary to perform simulations of different perturbations of the original state, and also the variation range of the moderator density can be higher than using the former KINPAR methodology. This is important for simulating severe accidents in which the variables vary in a wide range. This new methodology is applied to the simulation of the turbine trip transient Benchmark in Peach Bottom NPP using the TRAC-BF1 code. The results of the transient simulation in TRAC-BF1 using the KINPAR methodology and the new methodology, SIMTAB-1D, are compared. (author)

  20. Two phase nonequilibrium heat transfer in the TRAC-PD2 code

    International Nuclear Information System (INIS)

    Mandell, D.A.; Liles, D.R.

    1980-01-01

    TRAC is a best-estimate, multidimensional, nonequilibrium computer code intended for the analysis of loss-of-coolant accidents (LOCA's) in light water reactors. TRAC-PD2 is the third, detailed, pressurized water reactor version of the code. The TRAC code is modular both by components and by function. That is, vessels, pipes, pumps, etc. can be coupled together in any manner in order to simulate a reactor or a particular experimental facility. Individual physical phenomena are also coded in separate subroutines. This paper discusses the wall to fluid heat transfer coefficient correlations, the interfacial heat transfer models, and presents results for several experimental facilities

  1. Implantation of TRAC-PF1 computer code of VAX-11/750

    International Nuclear Information System (INIS)

    Madeira, A.A.; Souza Gouvea, L. de; Galetti, M.R.S.

    1988-01-01

    The implantation of TRAC-PF1 code, IBM version, in VAX-11/750 is described. This work provides Reator Department with an advanced best-estimate tool to perform loss-of-coolant accident analysis. (Author) [pt

  2. The application of the TRAC-PD2 code in the CANON experiment

    International Nuclear Information System (INIS)

    Neves Conti, T. das; Freitas, R.L.

    1991-09-01

    The TRAC code (Transient Reactor Analysis Code), developed in the Los Alamos National Laboratory, is used to accident analysis in light water reactor. The TRAC-PD2 version, used in this paper, has a refined dynamic flow model for two fluids, which is based on the conservation equations of mass, momentum and energy for liquid and vapor, allowing then a mechanical and thermal unbalance between phases. This paper presents a comparison of the TRAC-PD2 code with the CANON experiment, which simulates a Loss of Coolant Accident (LOCA) by depressurizing a horizontal tube filled with water at different temperatures. The experiment consists in a instantaneous rupture in one of the tube's edge, taking measures of pressure and void fraction during the transient. The TRAC-PD2 code results are in a good agreement with the pressure and void fraction evolution obtained in the CANON experiment. (author)

  3. Evaluation of the Trac-PF1 code for simulating the Neptun reflooding experiment

    International Nuclear Information System (INIS)

    Pontedeiro, A.C.; Galetti, M.R.S.

    1991-01-01

    The present work presents an assessment of the TRAC-BF1 code using the results of the NEPTUN experiment which simulates the reflooding in a loss-of-coolant accident (LOCA) in a PWR. The NEPTUN experiment is composed of an array of electrically-heated tubes where the reflooding condition can be tested. Two types of tests results are presented and compared with the values obtained with the TRAC-BF1 code. From this comparison it is concluded that TRAC is suitable for verifying accident analysis. (author)

  4. One dimensional neutron kinetics in the TRAC-BF1 code

    International Nuclear Information System (INIS)

    Weaver, W.L. III; Wagner, K.C.

    1987-01-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

  5. Methodology, status, and plans for development and assessment of the TRAC code

    International Nuclear Information System (INIS)

    Boyack, B.E.; Nelson, R.A.; Jolly-Woodruff, S.

    1996-01-01

    The Transient Reactor Analysis Code (TRAC) is a state-of-the-art, best-estimate, transient system analysis computer code for analyzing geometrically complex multidimensional thermal hydraulic systems, primarily nuclear reactor power plants. TRAC is used by government and industry organizations for design and safety analysis, phenomenological studies, operational transient analysis, evaluating emergency operating procedures, simulator support and operator training, and for assessment of data involving basic experiments, separate effects tests, and plant operations. TRAC will calculate one- and three-dimensional (rectilinear and cylindrical coordinates) fluid flow involving gas, liquid, and mixture states. Although TRAC has many capabilities, it also has limitations. Some limitations arise from its implementation, dating from the 1970s. Rapid advances in hardware and software engineering highlight TRAC's inefficiencies; however, other limitations relate to the level of scientific knowledge regarding two-phase flow physics. These limitations will continue until such time as the fundamental understanding of two-phase flows is extended. Presently, several development activities are either in progress or soon to begin that will fundamentally improve TRAC. Foremost among these are reimplementation of the current TRAC data structures in Fortran 90 and the integrated development of closure packages for large-break loss-of-coolant accident applications

  6. LINK codes TRAC-BF1/PARCSv2.7 in LINUX without external communication interface

    International Nuclear Information System (INIS)

    Barrachina, T.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Verdu, G.; Concejal, A.; Solar, A.

    2014-01-01

    The TRAC-BF1 code is still widely used by the nuclear industry for safety analysis. The plant models developed using this code are highly validated, so it is advisable to continue improving this code before migrating to another completely different code. The coupling with the NRC neutronic code PARCSv2.7 increases the simulation capabilities in transients in which the power distribution plays an important role. In this paper, the procedure for the coupling of TRAC-BF1 and PARCSv2.7 codes without PVM and in Linux is presented. (Author)

  7. Implementation of JAERI's reflood model into TRAC-PF1/MOD1 code

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

    1993-02-01

    Selected physical models of REFLA code, that is a reflood analysis code developed at JAERI, were implemented into the TRAC-PF1/MOD1 code in order to improve the predictive capability of the TRAC-PF1/MOD1 code for the core thermal hydraulic behaviors during the reflood phase in a PWR LOCA. Through comparisons of physical models between both codes, (1) Murao-Iguchi void fraction correlation, (2) the drag coefficient correlation acting to drops, (3) the correlation for wall heat transfer coefficient in the film boiling regime, (4) the quench velocity correlation and (5) heat transfer correlations for the dispersed flow regime were selected from the REFLA code to be implemented into the TRAC-PF1/MOD1 code. A method for the transformation of the void fraction correlation to the equivalent interfacial friction model was developed and the effect of the transformation method on the stability of the solution was discussed. Through assessment calculation using data from CCTF (Cylindrical Core Test Facility) flat power test, it was confirmed that the predictive capability of the TRAC code for the core thermal hydraulic behaviors during the reflood can be improved by the implementation of selected physical models of the REFLA code. Several user guidelines for the modified TRAC code were proposed based on the sensitivity studies on fluid cell number in the hydraulic calculation and on node number and effect of axial heat conduction in the heat conduction calculation of fuel rod. (author)

  8. TRAC User's Guide

    International Nuclear Information System (INIS)

    Boyack, B.E.; Stumpf, H.; Lime, J.F.

    1985-11-01

    This guide has been prepared to assist users in applying the Transient Reactor Analysis Code (TRAC). TRAC is an advanced best-estimate systems code for analyzing transients in thermal-hydraulic systems. The code is very general. Because it is general, efforts to model specific nuclear power plants or experimental facilities often present a challenge to the TRAC user. This guide has been written to assist first-time or intermediate users. It is specifically written for the TRAC version designated TRAC-PF1/MOD1. The TRAC User's Guide should be considered a companion document to the TRAC Code Manual; the user will need both documents to use TRAC effectively. 18 refs., 45 figs., 19 tabs

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

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

    1994-03-01

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

  10. Introduction and immigration of TRAC-PF1 code

    International Nuclear Information System (INIS)

    Yan Yuhua; Gao Zuying; Gao Cheng; Li Jincai

    1997-01-01

    TRAC-PF1 code performs best-estimate predictions of postulated accidents for pressurized light water reactors. It is one of the few system analyse codes which use two fluid model to treat two phase problems in nuclear system. In order to use this advanced software in China and make it possible to be run in different compute systems, IBM version of TRAC-PF1 code, imported from USA National Energy Software Center, is immigrated to CDC NOS/VE system and SUN workstation. The differences in computer languages from IBM 370 to CDC NOS/VE and to SUN workstation are modified properly. All the benchmark problems are calculated, and the results show that the immigration is successful

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  12. TRAC analysis of passive containment cooling system performance

    International Nuclear Information System (INIS)

    Arai, Kenji; Kataoka, Kazuyoshi; Nagasaka, Hideo

    1993-01-01

    A passive containment cooling system (PCCS) is a promising concept to improve the reliability of decay heat removal during an accident. Toshiba has carried out analytical studies for PCCS development in addition to experimental studies, using a best estimate thermal hydraulic computer code TRAC. In order to establish an analytical model for the PCCS performance analysis, it is necessary for the analytical model to be qualified against experimental results and thoroughly address the phenomena important for PCCS performance analysis. In this paper, the TRAC qualification for PCCS application is reported. A TRAC model has been verified against a drain line break simulation test conducted at the PCCS integral test facility, GIRAFFE. The result shows that the TRAC model can accurately predict the major system response and the PCCS performance in the drain line break test. In addition, the results of several sensitivity analyses, showing various points concerning the modeling in the PCCS performance analysis, have been reported. The analyses have been carried out for the SBWR and the analytical points are closely related to important phenomena which can affect PCCS performance

  13. Independent assessment of the TRAC-BD1/MOD1 computer code at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Wilson, G.E.; Charboneau, B.L.; Dallman, R.J.; Kullberg, C.M.; Wagner, K.C.; Wheatley, P.D.

    1984-01-01

    Under auspices of the United States Nuclear Regulatory Commission, their primary boiling water reactor safety analysis code (TRAC-BWR) is being assessed with simulations of a wide range of experimental data. The FY-1984 assessment activities were associated with the latest version (TRAC-BD1/MOD1) of this code. Typical results of the assessment studies are given. Conclusions formulated from these results are presented. These calculations relate to the overall applicability of the current code to safety analysis, and to future work which would further enhance the code's quality and ease of use

  14. TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1

    International Nuclear Information System (INIS)

    Meier, J.K.

    1983-01-01

    The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model

  15. TRAC-PD2 analysis of FLECHT experiments

    International Nuclear Information System (INIS)

    Bott, T.F.; Mandell, D.A.

    1981-01-01

    This report describes TRAC-PD2 calculations of FLECHT (Full Length Emergency Cooling Heat Transfer) tests 4831 and 17201. The calculations were performed as part of the TRAC-PD2 developmental assessment where the objective was to assess TRAC-PD2 reflood modeling under forced flooding conditions. Calculated and experimental values for peak fuel-rod clad temperature, clad quenching time, and rod bundle effluent rates are compared; and calculations with an approximate radiation heat-transfer model added to the basic TRAC-PD2 code are performed. Findings demonstrate the potential importance of surface-to-surface radiation heat transfer in these tests

  16. TRACMAB. A computer code to form part of the link between the codes TRAC and MABEL

    International Nuclear Information System (INIS)

    Newbon, S.

    1982-05-01

    This report describes the function of the link program TRACMAB and provides a guide for users. The program is required to convert the thermal disequilibrium data output by the transient code TRAC into equilibrium data in a format compatible with the input data required by the code CAIN which in turn produces input data for MABEL. (author)

  17. LINK codes TRAC-BF1/PARCSv2.7 in LINUX without external communication interface; Acoplamiento de los codigos TRAC-BF1/PARCSv2.7 en Linux sin interfaz externa de comunicacion

    Energy Technology Data Exchange (ETDEWEB)

    Barrachina, T.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Verdu, G.; Concejal, A.; Solar, A.

    2014-07-01

    The TRAC-BF1 code is still widely used by the nuclear industry for safety analysis. The plant models developed using this code are highly validated, so it is advisable to continue improving this code before migrating to another completely different code. The coupling with the NRC neutronic code PARCSv2.7 increases the simulation capabilities in transients in which the power distribution plays an important role. In this paper, the procedure for the coupling of TRAC-BF1 and PARCSv2.7 codes without PVM and in Linux is presented. (Author)

  18. TRAC methods and models

    International Nuclear Information System (INIS)

    Mahaffy, J.H.; Liles, D.R.; Bott, T.F.

    1981-01-01

    The numerical methods and physical models used in the Transient Reactor Analysis Code (TRAC) versions PD2 and PF1 are discussed. Particular emphasis is placed on TRAC-PF1, the version specifically designed to analyze small-break loss-of-coolant accidents

  19. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujits, R.K.

    1985-01-01

    A computer code (TRAC-PFI/MODI; denoted as TRAC) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the Once-Through Integral Systems (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and saturation, intermittent reactor coolant system circulation, boiler-condenser mode and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool- and auxiliary- feedwater initiated boiler-condenser mode heat transfer

  20. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujita, R.K.

    1985-01-01

    A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

  1. Improved gap conductance model for the TRAC code

    International Nuclear Information System (INIS)

    Hatch, S.W.; Mandell, D.A.

    1980-01-01

    The purpose of the present work, as indicated earlier, is to improve the present constant fuel clad spacing in TRAC-P1A without significantly increasing the computer costs. It is realized that the simple model proposed may not be accurate enough for some cases, but for the initial calculations made the DELTAR model improves the predictions over the constant Δr results of TRAC-P1A and the additional computing costs are negligible

  2. TRAC-PF1/MOD1 computer code

    International Nuclear Information System (INIS)

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

    1983-01-01

    TRAC-PF1 was designed to improve the ability of TRAC-PD2 to handle small-break LOCAs and other transients. TRAC-PF1 has all of the major improvements of TRAC-PD2 but, in addition, uses a full two-fluid model with two-step numerics in the one-dimensional components. The two-fluid model, in conjunction with a stratified-flow regime, handles countercurrent flow better than the drift-flux model previously used. The two-step numerics allow large time steps to be taken for slow transients. TRAC-PF1/MOD1 was designed to provide full balance-of-plant modeling capabilities. This required addition of a general capability for modeling plant control systems. The steam generator model was replaced to allow a wider variety of feedwater connections and better modeling of steam tube ruptures. A special turbine component also has been added, but new components were not required for adequate modeling of condensors, heaters, and pumps in the secondary system

  3. Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code

    International Nuclear Information System (INIS)

    Silva, J.D. da.

    1994-05-01

    In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs

  4. Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

    International Nuclear Information System (INIS)

    Beam, Tara M.; Ivanov, Kostadin N.; Baratta, Anthony J.; Finnemann, Herbert

    1999-01-01

    The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway

  5. Assessment of one dimensional reflood model in REFLA/TRAC code

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

    1993-12-01

    Post-test calculations for twelve selected SSRTF, SCTF and CCTF tests were performed to assess the predictive capability of the one-dimensional reflood model in the REFLA/TRAC code for core thermal behavior during the reflood in a PWR LOCA. Both core void fraction profile and clad temperature transients were predicted excellently by the REFLA/TRAC code including parameter effect of core inlet subcooling, core flooding rate, core configuration, core power, system pressure, initial clad temperature and so on. The peak clad temperature was predicted within an error of 50 K. Based on these assessment results, it is verified that the core thermal hydraulic behaviors during the reflood can be predicted excellently with the REFLA/TRAC code under various conditions where the reflood may occur in a PWR LOCA. (author)

  6. Application of the TRAC-PD2 code to the simulation of the CANON experiment

    International Nuclear Information System (INIS)

    Neves Conti, T. das; Freitas, R.L.

    1985-01-01

    A comparison between the TRAC -PD2 code calculations and results from the CANON experiment is presented. The CANON experiment simulates the loss of coolant accident through the depressurization of a horizontal tube containing water at different temperatures. The experiment consist of the instantaneous rupture at one end of the tubing and the corresponding pressure and void fraction measurements during the transient. The comparison shows that the TRAC-PD2 code predicts satisfactorily the pressure and void fraction evolution in the CANON experiment. (F.C.) [pt

  7. Post-test analysis of semiscale large-break test S-06-3 using TRAC-PF1

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1982-01-01

    The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because increased application of the code to long transients such as small-break LOCAs required a faster-running code version. Although developed for long transients, its performance on large-break transients is still important. This paper assesses the ability of TRAC-PF1 to predict large-break-LOCA Test S-06-3 conducted in the Semiscale Mod-1 facility

  8. Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

  9. MINI-TRAC code: a driver program for assessment of constitutive equations of two-fluid model

    International Nuclear Information System (INIS)

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

    1991-05-01

    MINI-TRAC code, a driver program for assessment of constitutive equations of two-fluid model, has been developed to perform assessment and improvement of constitutive equations of two-fluid model widely and efficiently. The MINI-TRAC code uses one-dimensional conservation equations for mass, momentum and energy based on the two-fluid model. The code can work on a personal computer because it can be operated with a core memory size less than 640 KB. The MINI-TRAC code includes constitutive equations of TRAC-PF1/MOD1 code, TRAC-BF1 code and RELAP5/MOD2 code. The code is modulated so that one can easily change constitutive equations to perform a test calculation. This report is a manual of the MINI-TRAC code. The basic equations, numerics, constitutive, equations included in the MINI-TRAC code will be described. The user's manual such as input description will be presented. The program structure and contents of main variables will also be mentioned in this report. (author)

  10. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

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

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  11. TRAC development at General Electric

    International Nuclear Information System (INIS)

    Andersen, J.G.M.; Shaug, J.C.; Shiralkar, B.S.

    1987-01-01

    TRAC is a computer code for transient analysis of light water reactors. The BWR version of TRAC has been developed as a result of a close cooperation between General Electric Company and Idaho National Engineering Laboratory. Up through 1985 the development work at General Electric was jointly funded by General Electric, the Nuclear Regulatory Commission and Electric Power Research Institute under the Refill-Reflood and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR) this work has led to the development of TRACBD1 and TRACBF1, while at GE, TRACB04 was the final product of the Refill-Reflood and FIST programs. TRAC development has continued at General Electric after the completion of these programs with the evolution of the TRACG code. The purpose of the paper is to describe this work. The TRAC development at General Electric can be divided into two main categories: extended benchmark capability and improved user convenience

  12. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

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

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. Calculations of Edwards' pipe blowdown tests using the code TRAC P1

    International Nuclear Information System (INIS)

    O'Mahoney, R.

    1979-05-01

    The paper describes the results obtained using the non-thermal equilibrium LOCA code TRAC-P1 for two of a series of Pipe Blowdown Tests. Comparisons are made with the experimental values and RELAP-UK Mark IV predictions. Some discrepancies between prediction and experiment are observed, and certain aspects of the model are considered to warrant possible further attention. (U.K.)

  15. TRAC analysis of design basis events for the accelerator production of tritium target/blanket

    International Nuclear Information System (INIS)

    Lin, J.C.; Elson, J.

    1997-01-01

    A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers

  16. Posttest analysis of MIST Test 330302 using TRAC-PF1/MOD1

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1992-09-01

    This report discusses a posttest analysis of Multi-Loop Integral System Test (MIST) 330302 which has been performed using TRAC-PF1/MOD1. This test was one of group performed in the MIST facility to investigate high-pressure injection (HPI)-power-operated relief valve (PORV) cooling, also known as feed-and-bleed cooling. In Test 330302, HPI cooling was delayed 20 min after opening and locking the PORV open to induce extensive system voiding. We have concluded that the TRAC-calculated results are in reasonable overall agreement with the data for Test 330302. All major trends and phenomena were correctly predicted. Differences observed between the measured and calculated results have been traced and related, in part, to deficiencies in our knowledge of the facility configuration and operation. We have identified two models forwhich additional review is appropriate. However, in general, the TRAC closure models and correlations appear to be adequate for the prediction of the phenomena expected to occur during feed-and-bleed transientsin the MIST facility. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications. Conclusions reached regarding use of the code to calculate similar phenomena in full-size plants (scaling implications) and regulatory implications of this work are also presented

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

    International Nuclear Information System (INIS)

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

    1990-08-01

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

  18. TRAC-PF1/MOD1 assessment at Los Alamos

    International Nuclear Information System (INIS)

    Knight, T.D.

    1984-01-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in pressurized water reactors (PWRs). Over the past several years, four distinct versions of the code have been released; each new version introduced improvements to the existing models and numerics and added new models to extend the applications of the code. The first goal of the code was to analyze large-break loss-of-coolant accidents (LOCAs), and the TRAC-P1A and TRAC-PD2 codes primarily addressed the large-break LOCA. (The TRAC-PD2/MOD1 code is essentially the same as the TRAC-PD2 code but it also includes a released set of error corrections.) The TRAC-PF1 code contained major changes to the models and trips and to the numerical methods. These modifications enhanced the computational speed of the code and improved the application to small-break LOCAs. The TRAC-PF1/MOD1 code, the latest released version, added improved steam-generator modeling, a turbine component, and a control system together with modified constitutive relations to model the balance of plant on the secondary side and to extend the applications to non-LOCA transients. The TRAC-PF1/MOD1 code also contains reasonably general reactor-kinetics modeling to facilitate the simulation of transients with delayed scram or without scram. 13 references, 24 figures

  19. TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

    International Nuclear Information System (INIS)

    Martin, R.P.

    1993-05-01

    Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

  20. Technical Requirements Analysis and Control Systems (TRACS) Initial Operating Capability (IOC) documentation

    Science.gov (United States)

    Hammond, Dana P.

    1991-01-01

    The Technical Requirements Analysis and Control Systems (TRACS) software package is described. TRACS offers supplemental tools for the analysis, control, and interchange of project requirements. This package provides the fundamental capability to analyze and control requirements, serves a focal point for project requirements, and integrates a system that supports efficient and consistent operations. TRACS uses relational data base technology (ORACLE) in a stand alone or in a distributed environment that can be used to coordinate the activities required to support a project through its entire life cycle. TRACS uses a set of keyword and mouse driven screens (HyperCard) which imposes adherence through a controlled user interface. The user interface provides an interactive capability to interrogate the data base and to display or print project requirement information. TRACS has a limited report capability, but can be extended with PostScript conventions.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. TRAC posttest calculations of Semiscale Test S-06-3

    International Nuclear Information System (INIS)

    Ireland, J.R.; Bleiweis, P.B.

    1980-01-01

    A comparison of Transient Reactor Analysis Code (TRAC) steady-state and transient results with Semiscale Test S-06-3 (US Standard Problem 8) experimental data is discussed. The TRAC model used employs fewer mesh cells than normal data comparison models so that TRAC's ability to obtain reasonable results with less computer time can be assessed. In general, the TRAC results are in good agreement with the data and the major phenomena found in the experiment are reproduced by the code with a substantial reduction in computing times

  3. Assessment of flooding in a best estimate thermal hydraulic code (WCOBRA/TRAC)

    International Nuclear Information System (INIS)

    Takeuchi, K.; Young, M.Y.

    1998-01-01

    The performance of WCOBRA/TRAC code in predicting the flooding, the counter-current flow limit, is evaluated in three geometries important to nuclear reactor loss-of-coolant accident evaluation; a vertical pipe, a perforated plate, and a downcomer annulus. These flow limits are computationally evaluated through transient conditions. The flooding in the vertical pipe is compared with the classical Wallis flooding limit. The flooding on the perforated plate is compared with the Northwestern flooding data correlation. The downcomer flooding in 1/15th and 1/5th scale model is compared with the Creare data. Finally, full scale downcomer flooding is compared with the UPTF test data. The prediction capability of the code for the flooding is found to be very good. (orig.)

  4. COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

    1987-01-01

    The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

  5. TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

    International Nuclear Information System (INIS)

    1978-05-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations

  6. TRAC calculations of a loss-of-coolant accident in a reactor scale model

    International Nuclear Information System (INIS)

    Pyun, J.J.

    1981-01-01

    The TRAC (Transient Reactor Analysis Code) is being developed at the Los Alamos National Laboratory as an advanced best-estimate computer program for analysis of postulated hypothetical accidents in pressurized water reactors. As a part of the TRAC developmental verification efforts, a TRAC posttest analysis of Semiscale Mod-3 Test S-07-6 was conducted. The results of this analysis show that the agreement between TRAC calculations and experimental data is not very good. In particular, TRAC does not predict the long term doncomer and core liquid level oscillations during the reflood phase

  7. Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code; Simulacao computacional dos experimentos: circulacao natural no CTE-150 e remolhamento na ITR utilizando o TRAC-PF1

    Energy Technology Data Exchange (ETDEWEB)

    Silva, J.D. da

    1994-05-01

    In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs.

  8. An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

    International Nuclear Information System (INIS)

    Hall, D.G.; Watkins, J.C.

    1987-01-01

    This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

  9. Independent assessment of TRAC-PD2 and RELAP5/MOD1 codes at BNL in FY 1981

    International Nuclear Information System (INIS)

    Saha, P.; Jo, J.H.; Neymotin, L.; Rohatgi, U.S.; Slovik, G.

    1982-12-01

    This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAP/MOD1 codes at Brookhaven National Laboratory (BNL) during Fiscal Year 1981. A large variety of separate-effects experiments dealing with (1) steady-state and transient critical flow, (2) level swell, (3) flooding and entrainment, (4) steady-state flow boiling, (5) integral economizer once-through steam generator (IEOTSG) performance, (6) bottom reflood, and (7) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2. In addition, the early part of an overcooling transient which occurred at the Rancho Seco nuclear power plant on March 20, 1978 was also computed with an updated version of TRAC-PD2. Three separate-effects tests dealing with (1) transient critical flow, (2) steady-state flow boiling, and (3) IEOTSG performance were also simulated with RELAP5/MOD1 code. Comparisons between the code predictions and the test data are presented

  10. TRAC-PF1 choked-flow model

    International Nuclear Information System (INIS)

    Sahota, M.S.; Lime, J.F.

    1983-01-01

    The two-phase, two-component choked-flow model implemented in the latest version of the Transient Reactor analysis Code (TRAC-PF1) was developed from first principles using the characteristic analysis approach. The subcooled choked-flow model in TRAC-PF1 is a modified form of the Burnell model. This paper discusses these choked-flow models and their implementation in TRAC-PF1. comparisons using the TRAC-PF1 choked-flow models are made with the Burnell model for subcooled flow and with the homogeneous-equilibrium model (HEM) for two-phae flow. These comparisons agree well under homogeneous conditions. Generally good agreements have been obtained between the TRAC-PF1 results from models using the choking criteria and those using a fine mesh (natural choking). Code-data comparisons between the separate-effects tests of the Marviken facility and the Edwards' blowdown experiment also are favorable. 10 figures

  11. TRAC-P validation test matrix. Revision 1.0

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. TRAC-P validation test matrix. Revision 1.0

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-05

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

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

    International Nuclear Information System (INIS)

    Schultz, R.R.

    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

  14. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

    2011-07-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  15. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  16. TRAC analysis of the Crystal River Unit-3 Plant transient of February 26, 1980

    International Nuclear Information System (INIS)

    Coddington, P.; Willcutt, G.J.E. Jr.

    1983-01-01

    This paper describes the application of the TRAC-PD2 and TRAC-PF1 codes to analyze the Crystal River transient. The PD2 and PF1 analyses used the three-dimensional and one-dimensional vessel models, respectively. Both calculations predicted the plant depressurization caused by the open PORV and the subsequent repressurization caused by closing the PORV and continuing high-pressure injection flow. Also, natural circulation was calculated in loop B following reestablishment of feedwater to the loop-B steam generator. After system repressurization, the codes calculated that pressure was relieved through the safety valves, and an intermittent flow occurred in loop A because of high-pressure-injection-driven density variations

  17. An HMS/TRAC analysis of a high-level radioactive waste tank

    International Nuclear Information System (INIS)

    Travis, J.R.; Nichols, B.D.; Spore, J.W.; Wilson, T.L.

    1991-01-01

    It has been observed that a high-level radioactive waste tank generates quantities of hydrogen and nitrous oxide mixtures that are potentially well within flammability limits. These gases are produced from chemical and nuclear decay reactions in a slurry of radioactive waste material. The slurry is covered by a thick crust composed of sodium nitrate and nitrite salts. Significant amounts of the combustible and reactant gases are produced over a 3- to 4-month period before the crust ruptures and the gases are vented into the air cover gas space above the crust. Postulating an ignition of the hydrogen/nitrous oxide/air mixture after this venting into the cover gas, we have calculated the pressure and temperature loading on the double-walled waste tank with the three-dimensional, time-dependent fluid dynamics coupled with chemical kinetics HMS (Hydrogen Mixing Studies) computer code. The waste tank has a ventilation system designed to maintain a slight negative gage pressure during steady-state operation. We have modeled the ventilation system with TRAC (the Transient Reactor Analysis Code), and we have coupled these two best-estimate accident analysis tools to provide the ventilation response to pressure and temperatures generated by the hydrogen burn. Significant pressures are produced by this event, and the threat to the tank's integrity currently is being evaluated. 3 refs., 4 figs

  18. Translation of the model plant of the CN code TRAC-BF1 Cofrentes of a SNAP-TRACE

    International Nuclear Information System (INIS)

    Escriva, A.; Munuz-Cobo, J. L.; Concejal, A.; Melara, J.; Albendea, M.

    2012-01-01

    It aims to develop a three-dimensional model of the CN Cofrentes whose consistent results Compared with those in current use programs (TRAC-BFl, RETRAN) validated with data of the plant. This comparison should be done globally and that you can not carry a compensation of errors. To check the correct translation of the results obtained have been compared with TRACE and the programs currently in use and the relevant adjustments have been made, taking into account that both the correlations and models are different codes. During the completion of this work we have detected several errors that must be corrected in future versions of these tools.

  19. Modelling of coupled self-actuating safety, relief and damped check valve systems with the codes TRAC-PF1 and ROLAST

    International Nuclear Information System (INIS)

    Neumann, U.; Puzalowski, R.; Grimm, I.

    1985-01-01

    Numerical valve models for simulation of selfactuating safety valves and damped check valves are introduced for the computer programs TRAC-PF1 and ROLAST. As examples of application post-test calculations and stability analysis are given. (orig.)

  20. Comparison of TRAC calculations with experimental data

    International Nuclear Information System (INIS)

    Jackson, J.F.; Vigil, J.C.

    1980-01-01

    TRAC is an advanced best-estimate computer code for analyzing postulated accidents in light water reactors. This paper gives a brief description of the code followed by comparisons of TRAC calculations with data from a variety of separate-effects, system-effects, and integral experiments. Based on these comparisons, the capabilities and limitations of the early versions of TRAC are evaluated

  1. TRAC code assessment using data from SCTF Core-III, a large-scale 2D/3D facility

    International Nuclear Information System (INIS)

    Boyack, B.E.; Shire, P.R.; Harmony, S.C.; Rhee, G.

    1988-01-01

    Nine tests from the SCTF Core-III configuration have been analyzed using TRAC-PF1/MOD1. The objectives of these assessment activities were to obtain a better understanding of the phenomena occurring during the refill and reflood phases of a large-break loss-of-coolant accident, to determine the accuracy to which key parameters are calculated, and to identify deficiencies in key code correlations and models that provide closure for the differential equations defining thermal-hydraulic phenomena in pressurized water reactors. Overall, the agreement between calculated and measured values of peak cladding temperature is reasonable. In addition, TRAC adequately predicts many of the trends observed in both the integral effect and separate effect tests conducted in SCTF Core-III. The importance of assessment activities that consider potential contributors to discrepancies between the measured and calculated results arising from three sources are described as those related to (1) knowledge about the facility configuration and operation, (2) facility modeling for code input, and (3) deficiencies in code correlations and models. An example is provided. 8 refs., 7 figs., 2 tabs

  2. Improvement of neutron kinetics module in TRAC-BF1code: one-dimensional nodal collocation method

    Energy Technology Data Exchange (ETDEWEB)

    Jambrina, Ana; Barrachina, Teresa; Miro, Rafael; Verdu, Gumersindo, E-mail: ajambrina@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidade Politecnica de Valencia (UPV), Valencia (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S.L., Madrid (Spain); Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion S.A.U., Madrid (Spain)

    2013-07-01

    The TRAC-BF1 one-dimensional kinetic model is a formulation of the neutron diffusion equation in the two energy groups' approximation, based on the analytical nodal method (ANM). The advantage compared with a zero-dimensional kinetic model is that the axial power profile may vary with time due to thermal-hydraulic parameter changes and/or actions of the control systems but at has the disadvantages that in unusual situations it fails to converge. The nodal collocation method developed for the neutron diffusion equation and applied to the kinetics resolution of TRAC-BF1 thermal-hydraulics, is an adaptation of the traditional collocation methods for the discretization of partial differential equations, based on the development of the solution as a linear combination of analytical functions. It has chosen to use a nodal collocation method based on a development of Legendre polynomials of neutron fluxes in each cell. The qualification is carried out by the analysis of the turbine trip transient from the NEA benchmark in Peach Bottom NPP using both the original 1D kinetics implemented in TRAC-BF1 and the 1D nodal collocation method. (author)

  3. TRAC development and assessment status

    International Nuclear Information System (INIS)

    Vigil, J.C.; Knight, T.D.

    1981-01-01

    TRAC is being developed at the Los Alamos National Laboratory to provide an advanced systems code for light-water reactor accident analysis. The released TRAC versions (P1, P1A, and PD2) were intended primarily as benchmark codes for large-break loss-of-coolant accidents but PD2 has been applied successfully to TMI-type transients and other small-break transients. A fast-running version, PFl, is currently under development to address more efficiently and accurately these types of transients. All of the released versions have been subjected to testing against separate-effects, system-effects, and integral experiments covering a wide range of scales. Assessment results indicate that PD2 does a credible job overall; needed improvements are being addressed in PFl and in modifications to PD2

  4. TRAC-PF1 code assessment using OECD LOFT LP-FP-1 experiment

    International Nuclear Information System (INIS)

    Barbero, F.J.

    1992-04-01

    This report assesses thermal-hydraulic aspects of LOFT LP-FP-1 experiment making use of TRAC-PF1/MOD1. LP-FP-1 experiment studies the system thermal-hydraulic and core thermal response for initial and boundary conditions similar to a large-break design basis LOCA leading to fission product release from the fuel cladding gap region. It also assesses the fission product retention effectiveness of the PWR-ECCS in best estimate conditions

  5. Improved timestep-size diagnostic edits for TRAC-P

    International Nuclear Information System (INIS)

    Giguere, P.T.

    1996-04-01

    Improvements have been made to the timestep-size selection logic diagnostic edits of the Transient Reactor Analysis Code (TRAC), specifically to the TRAC-P version. These include both a precise account of the reason for the selection for individual timesteps and thermal-hydraulic information on mesh cells that control the timestep size. The new edits can be specified by user input as a range of timestep numbers, problem time, or both. A description of the current timestep controls in effect in TRAC-P is also given

  6. Improved timestep-size diagnostic edits for TRAC-P

    Energy Technology Data Exchange (ETDEWEB)

    Giguere, P.T. [Los Alamos National Lab., NM (United States). Technology and Safety Assessment Div.

    1996-04-01

    Improvements have been made to the timestep-size selection logic diagnostic edits of the Transient Reactor Analysis Code (TRAC), specifically to the TRAC-P version. These include both a precise account of the reason for the selection for individual timesteps and thermal-hydraulic information on mesh cells that control the timestep size. The new edits can be specified by user input as a range of timestep numbers, problem time, or both. A description of the current timestep controls in effect in TRAC-P is also given.

  7. TRAC analysis and support of Oconee-1 PTS studies

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1983-01-01

    This paper describes the overall pressurized thermal shock (PTS) program at Los Alamos with emphasis on TRAC-PF1 calculations of severe overcooling transients for the Oconee-1 pressurized water reactor (PWR). A summary of results for several calculations are presented for the Oconee-1 PWR along with detailed discussions of two of the most severe overcooling transients predicted [main steam-line break and turbine-bypass valve (TBV) failures]. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to a detailed model of the plant Integrated Control System (ICS). For the Oconee-1 main steam-line break transient, a minimum downcomer fluid temperature of approx. 405 K was predicted. For the TBV transient involving the failure of one bank of TBVs to close after initially opening following reactor and turbine trips, an extrapolated downcomer fluid temperature of approx. 365 K was estimated. The latter temperature is at the nil-ductility temperature (NDT) limit (approx. 365 K) for Oconee-1

  8. Software design implementation document for TRAC-M data structures

    Energy Technology Data Exchange (ETDEWEB)

    Jolly-Woodruff, S. [Ogden Environmental and Energy Services (United States); Mahaffy, J. [Pennsylvania State Univ., University Park, PA (United States); Giguere, P.; Dearing, J.; Boyack, B. [Los Alamos National Lab., NM (United States)

    1997-07-01

    The Transient Reactor Analysis Code (TRAC)-M system-wide and component data structures are to be reimplemented by using the new features of Fortran 90 (F90). There will be no changes to the conceptual design, data flow, or computational flow with respect to the current TRAC-P, except that readability, maintainability, and extensibility will be improved. However, the task described here is a basic step that does not meet all future needs of the code, especially regarding extensibility. TRAC-M will be fully functional and will produce null computational changes with respect to TRAC-P, Version 5.4.25; computational efficiency will not be degraded significantly. The existing component and functional modularity and possibilities for coarse-grained parallelism will be retained.

  9. Software design implementation document for TRAC-M data structures

    International Nuclear Information System (INIS)

    Jolly-Woodruff, S.; Mahaffy, J.; Giguere, P.; Dearing, J.; Boyack, B.

    1997-07-01

    The Transient Reactor Analysis Code (TRAC)-M system-wide and component data structures are to be reimplemented by using the new features of Fortran 90 (F90). There will be no changes to the conceptual design, data flow, or computational flow with respect to the current TRAC-P, except that readability, maintainability, and extensibility will be improved. However, the task described here is a basic step that does not meet all future needs of the code, especially regarding extensibility. TRAC-M will be fully functional and will produce null computational changes with respect to TRAC-P, Version 5.4.25; computational efficiency will not be degraded significantly. The existing component and functional modularity and possibilities for coarse-grained parallelism will be retained

  10. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

  11. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

  12. Translation of the model plant of the CN code TRAC-BF1 Cofrentes of a SNAP-TRACE; Traduccion del modelo de planta de CN Cofrentes del codigo TRAC-BF1 a SNAP-TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Escriva, A.; Munuz-Cobo, J. L.; Concejal, A.; Melara, J.; Albendea, M.

    2012-07-01

    It aims to develop a three-dimensional model of the CN Cofrentes whose consistent results Compared with those in current use programs (TRAC-BFl, RETRAN) validated with data of the plant. This comparison should be done globally and that you can not carry a compensation of errors. To check the correct translation of the results obtained have been compared with TRACE and the programs currently in use and the relevant adjustments have been made, taking into account that both the correlations and models are different codes. During the completion of this work we have detected several errors that must be corrected in future versions of these tools.

  13. TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 x 15 generic four-loop Westinghouse nuclear power plant

    International Nuclear Information System (INIS)

    Spore, J.W.; Lin, J.C.; Schnurr, N.M.; White, J.R.; Cappiello, M.C.

    1992-01-01

    Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 x 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures

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

    International Nuclear Information System (INIS)

    Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

    1991-01-01

    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

  15. Modeling of High pressure core spray system (HPCS) with TRAC-BF1 code

    International Nuclear Information System (INIS)

    Angel M, E. Del.

    1993-01-01

    In this work we present a model of the HPCS system of Laguna Verde Nuclear Power Plant (CNLV) which consist of a condensate storage tank (TAC) a vertical surge pipe (TOS), implemented by the Comision Federal de Electricidad (CFE), and the suppression pool (PSP), as well as the associated piping and valves for the HPCS pump suction, to study the system under transient state conditions. The analysis results show that the implemented surge pipe, allows a normal HPCS pump start without automatic inadvertented transfer of the HPCS pump suction from the condensate storage tank to the suppression pool. We briefly mention some problems found during the stimulation and their solution, further we point out some deficiencies of the code for this type of studies. The present model can be used to stimulate other transients with only minor modifications of it. (Author)

  16. TRAC analysis of steam-generator overfill transients for TMI-1

    International Nuclear Information System (INIS)

    Bassett, B.

    1983-01-01

    A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

  17. TRAC Searchable Research Library

    Science.gov (United States)

    2016-05-01

    Relational Data Modeling (VRDM) computational paradigm. VRDM has the key attributes of being cloud available, using domain semantics for configured...Figure 1. Methodology for TRAC Searchable Research Library Development. ........................... 5 Figure 2. The conceptual model for the cloud ...TRAC Searchable Research Library project was initiated by TRAC- HQ to address a current capability gap in the TRAC organization. Currently TRAC does not

  18. Analysis of the FIST integral tests 4DBA1, 6SB2C and T1QUV with TRAC-BFl/v2001.2

    International Nuclear Information System (INIS)

    Analytis, G.Th.

    2004-01-01

    As part of the assessment of the frozen version of the PSU TRAC-BFl/v2001.2 and its qualification as a LB-LOCA and SB-LOCA code, in this work, we shall outline the comparisons between measurements and code predictions for three FIST tests: The LB-LOCA test 4DBA1, the SB-LOCA test 6SB2C and the failure to maintain water level test T1QUV. We shall study the effect of the number of axial levels in the active core as well as (in the case of the SB-LOCA test 6SB2C) the effect of the timing of the activation of the reflooding options/heat transfer package on the code predictions. Furthermore, we shall show that by using the upwinding option of some terms of the three-dimensional momentum equations, severe mass-error problems appearing in the analysis of the test T1QUV can be resolved. Generally, we shall show that although there are some differences between measurements and predictions, TRAC-BF1 captures quite well the overall behaviour of the LB-LOCA transient (depending on the number of axial nodes in the core) but underpredicts the rod surface temperatures of the SB-LOCA test 6SB2C

  19. Updated TRAC analysis of an 80% double-ended cold-leg break for the AP600 design

    International Nuclear Information System (INIS)

    Lime, J.F.; Boyack, B.E.

    1995-01-01

    An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) has been calculated for the Westinghouse AP600 advanced reactor design, The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse

  20. Modeling of High pressure core spray system (HPCS) with TRAC-BF1 code.; Modelacion del HPCS con el codigo TRAC-BF1.

    Energy Technology Data Exchange (ETDEWEB)

    Angel M, E Del

    1994-12-31

    In this work we present a model of the HPCS system of Laguna Verde Nuclear Power Plant (CNLV) which consist of a condensate storage tank (TAC) a vertical surge pipe (TOS), implemented by the Comision Federal de Electricidad (CFE), and the suppression pool (PSP), as well as the associated piping and valves for the HPCS pump suction, to study the system under transient state conditions. The analysis results show that the implemented surge pipe, allows a normal HPCS pump start without automatic inadvertented transfer of the HPCS pump suction from the condensate storage tank to the suppression pool. We briefly mention some problems found during the stimulation and their solution, further we point out some deficiencies of the code for this type of studies. The present model can be used to stimulate other transients with only minor modifications of it. (Author).

  1. Development of Monte Carlo input code for proton, alpha and heavy ion microdosimetric trac structure simulations

    International Nuclear Information System (INIS)

    Douglass, M.; Bezak, E.

    2010-01-01

    Full text: Radiobiology science is important for cancer treatment as it improves our understanding of radiation induced cell death. Monte Carlo simulations playa crucial role in developing improved knowledge of cellular processes. By model Ii ng the cell response to radiation damage and verifying with experimental data, understanding of cell death through direct radiation hits and bystander effects can be obtained. A Monte Carlo input code was developed using 'Geant4' to simulate cellular level radiation interactions. A physics list which enables physically accurate interactions of heavy ions to energies below 100 e V was implemented. A simple biological cell model was also implemented. Each cell consists of three concentric spheres representing the nucleus, cytoplasm and the membrane. This will enable all critical cell death channels to be investigated (i.e. membrane damage, nucleus/DNA). The current simulation has the ability to predict the positions of ionization events within the individual cell components on I micron scale. We have developed a Geant4 simulation for investigation of radiation damage to cells on sub-cellular scale (∼I micron). This code currently allows the positions of the ionisation events within the individual components of the cell enabling a more complete picture of cell death to be developed. The next stage will include expansion of the code to utilise non-regular cell lattice. (author)

  2. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

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

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  3. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  4. Input data preparation and simulation of the second standard problem of IAEA using the Trac/PF1 code

    International Nuclear Information System (INIS)

    Madeira, A.A.; Pontedeiro, A.C.; Silva Galetti, M.R. da; Borges, R.C.

    1989-10-01

    The second Standard Problem sponsored by IAEA consists in the simulation of a small LOCA located in the downcomer of a PMK-NVH integral test facility, which models WWER/440 type reactor. This report presents input data preparation and comparison between TRAC-PF1 results and experimental measurements. (author) [pt

  5. Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

    International Nuclear Information System (INIS)

    Henninger, R.J.; Boyack, B.E.

    1986-01-01

    A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

  6. Resuspension parameters for TRAC dispersion model

    International Nuclear Information System (INIS)

    Langer, G.

    1987-01-01

    Resuspension factors for the wind erosion of soil contaminated with plutonium are necessary to run the Rocky Flats Plant Terrain Responsive Atmospheric Code (TRAC). The model predicts the dispersion and resulting population dose due to accidental plutonium releases

  7. TRAC-PD2 modeling of LOFT and PWR small cold-leg breaks

    International Nuclear Information System (INIS)

    Knight, T.D.; Willcutt, G.J.E. Jr.; Lime, J.F.

    1981-01-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light-water reactors. TRAC-PD2, the latest publicly released version of the code, is currently being tested against small-break and other transients in experimental facilities; it is also being used to analyze postulated accidents in commercial power reactors. Calculated results for LOFT small-break experiments are compared to data, and the results from two small-break calculations for two different reactor systems are presented. It is concluded that TRAC-PD2 is useful for the analysis of cold-leg small-break accidents

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

    International Nuclear Information System (INIS)

    Sicilian, J.M.

    1977-06-01

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

  9. TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

    International Nuclear Information System (INIS)

    Lime, J.F.; Boyack, B.E.

    1996-01-01

    An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

  10. Presentation of geometries and transient results of TRAC-calculations

    International Nuclear Information System (INIS)

    Lutz, A.; Lang, U.; Ruehle, R.

    1985-02-01

    The computer code TRAC is used to analyze the transient behaviour of nuclear reactors. The input of a TRAC-Calculation, as well as the produced result files serve for the graphical presentation of the geometries and transient results. This supports the search for errors during input generation and the understanding of complex processes by dynamic presentation of calculational result in colour. (orig.) [de

  11. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    Ireland, J.R.; Boyack, B.E.

    1984-01-01

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

  12. Assessment of a new interfacial friction correlation in TRAC-BD1/MOD1

    International Nuclear Information System (INIS)

    Analytis, G.T.

    1986-01-01

    Analysis of a number of 5-bar (and one 1-bar) boiloff experiments in the electrically heated 33-rod bundle NEPTUN at the Swiss Federal Institute for Reactor Research by TRAC-BD1 version 12 and MOD1 has shown that the code consistently underpredicts the collapsed liquid level histories, hence predicting earlier critical heat fluxes and higher peak rod surface temperatures than the measurements showed. Moreover, recent work has demonstrated that these differences can be attributed to the bubbly/churn interfacial friction model in TRAC-BD1 (resulting in rather large interfacial drag) whose appropriateness for rod bundles is questionable; decreasing the interfacial drag resulted in excellent agreement between measurements and code predictions. Recent analysis of boiloff experiments with the French code CATHARE and with a bubbly/churn interfacial drag force f/sub i/ similar to the one of TRAC-BD1 has also resulted in the underprediction of collapsed liquid level histories. In this work, it was shown that the usual vapor drift velocity correlations (through which f/sub i/ is derived) for this flow regime developed for tubes are not appropriate for rod bundles. Moreover, a new f/sub i/ correlation for bubbly/churn flow in rod bundles was developed based on the Froude number. The authors have modified this correlation slightly, implemented it in TRAC-BD1/MOD1, and reanalyzed most of the boiloff experiments in NEPTUN

  13. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200% double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding temperature excursion was terminated at 175 s when complete core quenching occurred. Accumulator flows were initiated at 10 s when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching

  14. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

  15. Special small-break applications with TRAC

    International Nuclear Information System (INIS)

    Dobranich, D.; DeMuth, N.S.; Henninger, R.J.; Burns, R.D. III.

    1981-01-01

    Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical simulations. Also, the effects of steam generator recirculation, steam generator tube rupture, Emergency Core Cooling (ECC) injection and reactivity feedback on the course and consequences of these transients are investigated. These models allow reliable predictions of accident signatures that can help determine the adequacy of equipment and procedures at nuclear power plants to prevent and to control severe accidents

  16. Application of a generalized interface module to the coupling of PARCS with both RELAP5 and TRAC-M

    Energy Technology Data Exchange (ETDEWEB)

    Barber, D.A.; Wang, W. [SCIENTECH, Inc. (United States); Miller, R.M.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States); Joo, H.G. [Korean Atomic Energy Research Inst., Seoul (Korea, Republic of); Mousseau, V.A. [Los Alamos National Lab., NM (United States); Ebert, D.E. [Nuclear Regulatory Commission, Washington, DC (United States)

    1999-04-01

    In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem.

  17. Application of a generalized interface module to the coupling of PARCS with both RELAPS and TRAC-M

    International Nuclear Information System (INIS)

    Barber, D.A.; Wang, W.; Miller, R.M.; Downar, T.J.; Joo, H.G.; Mousseau, V.A.; Ebert, D.E.

    1999-01-01

    In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem

  18. Post-test analysis of semiscale tests S-UT-6 and S-UT-7 using TRAC PF1

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1983-01-01

    A posttest study of Semiscale Tests S-UT-6 and S-UT-7 has been completed to assess TRAC-PFl predictions of pressurized water-reactor (PWR) small-break transients. The comparisons of the TRAC calculations and experimental results show that the correct qualitative influence of upper-head injection (UHI) was predicted. The major phenomenological difference predicted was the mode of core voiding. The data show a slow boiloff from the top of the core resulting in a dryout near the top of the core only. TRAC predicted a more extensive voiding with fluid forced from the bottom of the core by a pressure increase in the upper vessel plenum. The pressure increase was the primary consequence of a failure to predict a complete clearance of the seal in the intact-loop pump-suction upflow leg. Further review of the interphasic drag correlations, entrainment correlations, and critical-flow model is recommended. 20 figures

  19. TRAC-PF1 MOD1 post test calculations of the OECD LOFT Experiment LP-SB-1

    International Nuclear Information System (INIS)

    Allen, E.J.

    1990-04-01

    Analysis of the small, hot leg break, OECD LOFT Experiment LP-SB-1. using the ''best-estimate'' computer code TRAC-PF1/MOD1 is presented. Descriptions of the LOFT facility and the LP-SB-1 experiment are given and development of the TRAC-PF1/MOD1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experiment was initiated, and the specification of experimental boundary conditions are outlined. 24 refs., 66 figs., 12 tabs

  20. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

    2007-07-01

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

  1. Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

    International Nuclear Information System (INIS)

    Todorova, N.K.; Ivanov, K.N.

    2003-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement

  2. Evaluation of fuel-temperature feedback mechanisms in TRAC-PF1/MOD2/NESTLE

    International Nuclear Information System (INIS)

    Knepper, Paula L.; Feltus, Madeline; Hochreiter, L.E.; Ivanov, Kostadin

    1999-01-01

    Coupled spatial kinetics and thermal-hydraulics system codes provide a means to model transient nuclear reactor behavior more accurately. Transients marked by strong perturbations, both with thermal-hydraulics and neutronics, such as a control-rod ejection or a main steam-line break, are especially of interest. It is now feasible to model complex reactor behavior with a coupled thermal-hydraulics and spatial kinetics code that provides a means to forecast safety margins. Recently, the Transient Reactor Analysis Code (TRAC)-PF1/MOD2, Version 5.4.25, was coupled with the NESTLE code. This coupled code (TRAC-PF1/MOD2/NESTLE) is used to examine effective fuel-temperature models. The Electric Power Research Institute (EPRI) rod-ejection benchmark was analyzed to evaluate the influence of effective fuel temperature. The rod-ejection transient tests only the fuel-rod, heat-conduction coupling. The coolant thermal-hydraulic coupling is not tested because of the speed of the transient. The neutronics solution changes extremely rapidly, whereas the convective heat transfer at the fuel surface requires more time to influence the coolant temperature of the system. The need to model the response of the system coolant temperature is not crucial in this analysis. The influence of the effective fuel temperature is the key component of this study. Various models were examined using the coupled code to calculate effective fuel temperatures. The influence of different, effective fuel-temperature models on the coupled-code results is studied. Three effective fuel-temperature models are examined: (l) volume average effective fuel temperature, (2) the effective fuel-temperature model suggested by the Office of Economic Cooperation and Development (OECD) rod-ejection benchmark, and (3) the NESTLE effective fuel-temperature model. A discussion is provided describing the effective fuel-temperature models examined in TRAC-PF1/MOD2/NESTLE and the influence of effective fuel temperature in

  3. Quantifying reactor safety margins: Application of CSAU [Code Scalability, Applicability and Uncertainty] methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

    International Nuclear Information System (INIS)

    Wulff, W.; Boyack, B.E.; Duffey, R.B.

    1988-01-01

    Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs

  4. Improved coupling of the conduction and flow equations in TRAC

    International Nuclear Information System (INIS)

    Addessio, F.L.

    1981-01-01

    Recent nuclear-reactor-systems modeling efforts have been directed toward the development of computer codes capable of simulating transients in short computational times. For this reason, a stability enhancing two-stem method (SETS) has been applied to the two-phase flow equations in the Transient Reactor Analysis Code (TRAC) allowing the Courant limit to be violated. Unfortunately, the coupling between the wall conduction equation and the fluid-dynamics equations is performed semi-implicitly, that is, the wall-heat transfer term, is evaluated using old-time heat-transfer coefficients and wall temperatures and new-time coolant temperatures. This coupling may lead to numerical instabilities at large time steps because of large variations in the heat-transfer coefficient in certain regimes of the boiling curve. Consequently, simply using new-time wall temperatures is not sufficient. A technique that also incorporates new-time heat-transfer coefficients must be used

  5. Comparisons of TRAC-PD2 calculations with Semiscale Mod-3 small-break tests

    International Nuclear Information System (INIS)

    Gilbert, J.S.; Sahota, M.S.; Boyack, B.E.; Booker, C.P.; Meier, J.K.

    1981-01-01

    Five experiments conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory (INEL) were calculated using the latest released version of the Transient Reactor Analysis Code (TRAC-PD2). The results were used to assess TRAC-PD2 predictions of thermal-hydraulic phenomena and the effects of pump operation on system response during slow transients. Tests S-SB-P1, S-SB-P2, and S-SB-P7 simulated equivalent 2.5% communicative cold-leg breaks for early pump-trip (pumps-off), intermediate pump-trip (pumps-on), and late pump-trip (pumps-on) operation, respectively. Tests S-SB-P3 and S-SB-P4 simulated equivalent 2.5% communicative hot-leg breaks for pumps-off and pumps-on operation, respectively. Parameters examined in the study included primary system mass distribution, mass inventory, and void fraction distribution

  6. Three-dimensional two-fluid numerical treatment of a reactor vessel in TRAC

    International Nuclear Information System (INIS)

    Liles, D.R.

    1979-01-01

    A three-dimensional two-fluid finite difference model has been used in TRAC (Transient Reactor Analysis Code) to represent a pressurized water reactor vessel. Mesh cells may be blocked off completely to represent large flow obstructions such as downcomer walls. The hydrodynamic volumes and flow areas may also be reduced in order to provide a porous matrix simulation of smaller scale strucuture. The finite difference equations are semi-implicit so that stability time scales are associated with material movement and not wave propagation. The block matrix structure is reduced during the implicit pass to a single element seven stripe system which is easily solved iteratively. This procedure has successfully performed numerous simulations of both full sized reactor accidents and smaller scale experments. It has proven to be a useful feature of the TRAC effort

  7. TRAC Innovative Visualization Techniques

    Science.gov (United States)

    2016-11-14

    5Benjamin Jotham Fry. “Computational information design”. PhD thesis . Massachusetts Institute of Technology, 2004, p. 39. 6Ibid., p. 39. 7Tamara Munzner...central capacity limit to the simultaneous storage of visual and auditory arrays in working memory.” In: Journal of Experimental Psychology : General...GAT scores. 18 August 2016 TRAC Brief Template (AE Method) 2 1Q 2Q 3Q SpiritualService 4Q Rank Gender Spiritual LowVery Low High Initial Model Final

  8. Development of computer aided engineering system for TRAC applications

    International Nuclear Information System (INIS)

    Arai, Kenji; Itoya, Seihiro; Uematsu, Hitoshi; Tsunoyama, Shigeaki

    1990-01-01

    An advanced best estimate computer program for nuclear reactor transient analysis, TRAC has been extensively used to carry out various thermal hydraulic calculations in the nuclear engineering field, because of its versatility. To perform efficiently a wide variety of TRAC calculation, the efficient utilization of computers and the convenient environment for input and output processing is necessary. We have applied a computer network comprising a super-computer, engineering work stations and personal computers to TRAC calculations and have assigned the appropriate functions to each computer. We have also been developing an interactive graphics system for input and output processing on an EWS. This hardware and software environment can improve the effectiveness of TRAC utilization for various thermal hydraulic calculations. (author)

  9. Heat Transfer treatment in computer codes for safety analysis

    International Nuclear Information System (INIS)

    Jerele, A.; Gregoric, M.

    1984-01-01

    Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)

  10. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

    Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

  11. Assessment of TRAC-BF1 1D reflood model with CCTF and SCTF data

    International Nuclear Information System (INIS)

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

    1993-03-01

    Post test calculations for six selected Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF) tests were performed to assess the core thermal hydraulic models of the TRAC-BF1 code during the reflood in a PWR LOCA. A special version of the TRAC code was developed at JAERI by implementing the constitutive package of the TRAC-BF1 code into the TRAC-PF1 code for this assessment. The TRAC-BF1 model predicted well the void fraction at either bottom or top part of the core and overpredicted the void fraction at the center part of the core in the CCTF and SCTF tests performed under so-called licensing conditions. The TRAC-BF1 model overpredicted the clad temperatures at the center part of the core. The TRAC-BF1 model predicted a jump of void fraction where the flow pattern transition between the bubbly/slug flow and the annular/dispersed flow regimes occurred. The jump caused the water mass flow rate to be unstable and resulted in the overprediction of the void fraction at the center part of the core. It was also found that the TRAC-BF1 film boiling model underestimated the heat transfer coefficient in the vicinity of the quench front and caused the quench front propagation to be delayed. These assessment results suggest the following areas should be improved in future to apply the TRAC-BF1 code to the reflood in a PWR LOCA: (1) Core hydraulic model where flow pattern transition occurs, (2) Core heat transfer model in the film boiling regime, especially for the dependence on the distance from the quench front. (author)

  12. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

  13. TRAC-PF1/MOD2 status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Steinke, R.G.; Nelson, R.A.; Cappiello, M.W.; Jenks, R.

    1989-01-01

    The development of the TRAC-PF1/MOD1 code was completed in July 1988 with the release of Version 14.4. A TRAC-PF1/MOD2 code development plan addresses code deficiencies identified in the MOD1 code in order to provide an accurate and defensible tool that can be used to simulate large-break loss-of-coolant accidents (LOCAs), small-break LOCAs, and operational transients. The MOD2 code development plan is an international cooperative effort that includes contributions from Los Alamos National Laboratory, Idaho National Engineering Laboratory (INEL), Japanese Atomic Energy Research Institute (JAERI), Cray Research, Central Electricity Generating Board (CEGB), and United Kingdom Atomic Energy Authority (UKAEA)

  14. System calculations related to the accident at Three-Mile Island using TRAC

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1980-01-01

    The Three Mile Island nuclear plant (Unit 2) was modeled using the Transient Reactor Analysis Code (TRAC-P1A) and a base case calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. In addition to the base case calculation, several parametric calculations were performed in which a single hypothetical change was made in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident. Some of the important system parameter comparisons for the base case as well as some of the parametric case results are presented

  15. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  16. One-dimensional nodal neutronics routines for the TRAC-BD1 thermal-hydraulics program

    International Nuclear Information System (INIS)

    Nigg, D.W.

    1983-09-01

    Nuclear reactor core transient neutronic behavior is currently modeled in the TRAC-BD1 code using a point-reactor kinetics formulation. This report describes a set of subroutines based on the Analytic Nodal Method that were written to provide TRAC-BD1 with a one-dimensional space-dependent neutronics capability. Use of the routines is illustrated with several test problems. The results of these problems show that the Analytic Nodal neutronics routines have desirable accuracy and computing time characteristics and should be a useful addition to TRAC-BD1

  17. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  18. Uncertainties in modelling and scaling of critical flows and pump model in TRAC-PF1/MOD1

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Yu, Wen-Shi.

    1987-01-01

    The USNRC has established a Code Scalability, Applicability and Uncertainty (CSAU) evaluation methodology to quantify the uncertainty in the prediction of safety parameters by the best estimate codes. These codes can then be applied to evaluate the Emergency Core Cooling System (ECCS). The TRAC-PF1/MOD1 version was selected as the first code to undergo the CSAU analysis for LBLOCA applications. It was established through this methodology that break flow and pump models are among the top ranked models in the code affecting the peak clad temperature (PCT) prediction for LBLOCA. The break flow model bias or discrepancy and the uncertainty were determined by modelling the test section near the break for 12 Marviken tests. It was observed that the TRAC-PF1/MOD1 code consistently underpredicts the break flow rate and that the prediction improved with increasing pipe length (larger L/D). This is true for both subcooled and two-phase critical flows. A pump model was developed from Westinghouse (1/3 scale) data. The data represent the largest available test pump relevant to Westinghouse PWRs. It was then shown through the analysis of CE and CREARE pump data that larger pumps degrade less and also that pumps degrade less at higher pressures. Since the model developed here is based on the 1/3 scale pump and on low pressure data, it is conservative and will overpredict the degradation when applied to PWRs

  19. Nuclear Plant Analyzer: an interactive TRAC/RELAP Power-Plant Simulation Program

    International Nuclear Information System (INIS)

    Steinke, R.; Booker, C.; Giguere, P.; Liles, D.; Mahaffy, J.; Turner, M.; Wiley, R.

    1984-01-01

    The Nuclear Plant Analyzer (NPA) is a computer-software interface for executing the TRAC or RELAP5 power-plant systems codes. The NPA is designed to use advanced supercomputers, long-distance data communications, and a remote workstation terminal with interactive computer graphics to analyze power-plant thermal-hydraulic behavior. The NPA interface simplifies the running of these codes through automated procedures and dialog interaction. User understanding of simulated-plant behavior is enhanced through graphics displays of calculational results. These results are displayed concurrently with the calculation. The user has the capability to override the plant's modeled control system with hardware adjustment commands. This gives the NPA the utility of a simulator, and at the same time, the accuracy of an advanced, best-estimate, power-plant systems code for plant operation and safety analysis

  20. Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)

    2014-05-15

    The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

  1. Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

    International Nuclear Information System (INIS)

    Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong

    2014-01-01

    The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

  2. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  3. AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

    International Nuclear Information System (INIS)

    Knight, T.D.

    1996-07-01

    The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA

  4. Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7

    International Nuclear Information System (INIS)

    Sahota, M.S.

    1982-01-01

    Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

  5. TRAC-Monterey FY16 Work Program Development and Report of Research Elicitation

    Science.gov (United States)

    2016-01-01

    any changes to priorities or additional projects that require immediate research. Work Program; Research Elicitation Unclassified UU UU UU UU 35 MAJ...conduct analysis for the Army. 1 Marks, Chris, Nesbitt, Peter. TRAC FY14 Research Requirements Elicitation . Technical Report TRAC-M-TM-13-059. 700 Dyer... Requirements Elicitation Interviews Interview Guide: 1. Describe a research requirement in the areas of topics, techniques, and methodologies. 2

  6. TRAC analyses of severe overcooling transients for the Oconee-1 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ireland, J R [comp.

    1985-05-01

    This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

  7. TRAC analyses of severe overcooling transients for the Oconee-1 PWR

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1985-05-01

    This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs

  8. Improvements to TRAC models of condensing stratified flow. Pt. 1

    International Nuclear Information System (INIS)

    Zhang, Q.; Leslie, D.C.

    1991-12-01

    Direct contact condensation in stratified flow is an important phenomenon in LOCA analyses. In this report, the TRAC interfacial heat transfer model for stratified condensing flow has been assessed against the Bankoff experiments. A rectangular channel option has been added to the code to represent the experimental geometry. In almost all cases the TRAC heat transfer coefficient (HTC) over-predicts the condensation rates and in some cases it is so high that the predicted steam is sucked in from the normal outlet in order to conserve mass. Based on their cocurrent and countercurrent condensing flow experiments, Bankoff and his students (Lim 1981, Kim 1985) developed HTC models from the two cases. The replacement of the TRAC HTC with either of Bankoff's models greatly improves the predictions of condensation rates in the experiment with cocurrent condensing flow. However, the Bankoff HTC for countercurrent flow is preferable because it is based only on the local quantities rather than on the quantities averaged from the inlet. (author)

  9. Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1

    International Nuclear Information System (INIS)

    Castrillo, F.; Gomez, A.; Gallego, I.

    1993-06-01

    This report presents the results of the assessment of TRAC-BF1 (G1-J1) code with the model of C. N. Cofrentes for simulation of the transient originated by the manual trip of the main turbine. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 Mwt, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola, S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbine driven FW pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator

  10. Analysis of an ADS spurious opening event at a BWR/6 by means of the TRACE code

    International Nuclear Information System (INIS)

    Nikitin, Konstantin; Manera, Annalisa

    2011-01-01

    Highlights: → The spurious opening of 8 relief valves of the ADS system in a BWR/6 has been simulated. → The valves opening results in a fast depressurization and significant loads on the RPV internals. → This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the available plant data. - Abstract: The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.

  11. Safety analysis code input automation using the Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Kopp, H.; Leung, J.; Tajbakhsh, A.; Viles, F.

    1985-01-01

    The Nuclear Plant Data Bank (NPDB) is a computer-based system that organizes a nuclear power plant's technical data, providing mechanisms for data storage, retrieval, and computer-aided engineering analysis. It has the specific objective to describe thermohydraulic systems in order to support: rapid information retrieval and display, and thermohydraulic analysis modeling. The Nuclear Plant Data Bank (NPBD) system fully automates the storage and analysis based on this data. The system combines the benefits of a structured data base system and computer-aided modeling with links to large scale codes for engineering analysis. Emphasis on a friendly and very graphically oriented user interface facilitates both initial use and longer term efficiency. Specific features are: organization and storage of thermohydraulic data items, ease in locating specific data items, graphical and tabular display capabilities, interactive model construction, organization and display of model input parameters, input deck construction for TRAC and RELAP analysis programs, and traceability of plant data, user model assumptions, and codes used in the input deck construction process. The major accomplishments of this past year were the development of a RELAP model generation capability and the development of a CRAY version of the code

  12. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  13. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  14. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  15. 76 FR 32227 - Core Industries, Inc., DBA Star Trac and/or Unisen, Inc., DBA STAR TRAC and/or Trac Strength...

    Science.gov (United States)

    2011-06-03

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-75,192; TA-W-75,192A] Core Industries, Inc., DBA Star Trac and/or Unisen, Inc., DBA STAR TRAC and/or Trac Strength, Including On-Site Leased Workers From Aerotek, Helpmates, Mattson, and Empire Staffing, Irvine, CA; Core Industries, Inc., DBA Star Trac and/ar Unisen, Inc., DBA Star...

  16. Balance of plant modeling in TRAC-BD1/MOD1

    International Nuclear Information System (INIS)

    Weaver, W.L.; Giles, M.M.; Mohr, C.M.

    1983-01-01

    The mission of the TRAC-BD1/MOD1 code is to provide a best-estimate analysis capability for Boiling Water Reactor systems and related experimental facilities for the full range of accidents from large and small break Loss-of-Coolant accidents to operational transients including anticipated transients without scram (ATWS), for which point reactor kinetics is adequate (as a first approximation). Recent model developments allow a complete reactor system including the containment and the balance of plant to be modeled. This paper describes the balance of plant models and presents the results of a simulation of a loss-of-feedwater heater transients which was used to assess the performance of the balance of plant models

  17. Simulation of LOFT anticipated-transient experiments L6-1, L6-2, and L6-3 using TRAC-PF1/MOD1

    International Nuclear Information System (INIS)

    Sahota, M.S.

    1984-01-01

    Anticipated-transient experiments L6-1, L6-2, and L6-3, performed at the Loss-of-fluid Test (LOFT) facility, are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1/MOD1). The results are used to assess TRAC-PF1/MOD1 trip and control capabilities, and predictions of thermal-hydraulic phenomena during slow transients. Test L6-1 simulated a loss-of-stream load in a large pressurized-water reactor (PWR), and was initiated by closing the main steam-flow control valve (MSFCV) at its maximum rate, which reduced the heat removal from the secondary-coolant system and increased the primary-coolant system pressure that initiated a reactor scram. Test L6-2 simulated a loss-of-primary coolant flow in a large PWR, and was initiated by tripping the power to the primary-coolant pumps (PCPs) allowing the pumps to coast down. The reduced primary-coolant flow caused a reactor scram. Test L6-3 simulated an excessive-load increase incident in a large PWR, and was initiated by opening the MSFCV at its maximum rate, which increased the heat removal from the secondary-coolant system and decreased the primary-coolant system pressure that initiated a reactor scram. The TRAC calculations accurately predict most test events. The test data and the calculated results for most parameters of interest also agree well

  18. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States). Dept. of Nuclear Engineering; Wang, W. [SCIENTECH, Inc., Rockville, MD (United States); Mousseau, V.A.; Ebert, D.D. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    1999-03-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.

  19. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Mousseau, V.A.; Ebert, D.D.

    1999-01-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model

  20. Stability analysis by ERATO code

    International Nuclear Information System (INIS)

    Tsunematsu, Toshihide; Takeda, Tatsuoki; Matsuura, Toshihiko; Azumi, Masafumi; Kurita, Gen-ichi

    1979-12-01

    Problems in MHD stability calculations by ERATO code are described; which concern convergence property of results, equilibrium codes, and machine optimization of ERATO code. It is concluded that irregularity on a convergence curve is not due to a fault of the ERATO code itself but due to inappropriate choice of the equilibrium calculation meshes. Also described are a code to calculate an equilibrium as a quasi-inverse problem and a code to calculate an equilibrium as a result of a transport process. Optimization of the code with respect to I/O operations reduced both CPU time and I/O time considerably. With the FACOM230-75 APU/CPU multiprocessor system, the performance is about 6 times as high as with the FACOM230-75 CPU, showing the effectiveness of a vector processing computer for the kind of MHD computations. This report is a summary of the material presented at the ERATO workshop 1979(ORNL), supplemented with some details. (author)

  1. TRAC-PF1/MOD1 independent assessment: Semiscale Mod-2A intermediate break test S-IB-3

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1986-02-01

    The TRAC-PF1/MOD1 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various system codes to predict the detailed thermal/hydraulic response of light water reactors during accident and off-normal conditions. The TRAC code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, an intermediate break test (S-IB-3), performed at the Semiscale Mod-2A facility, has been analyzed. Using an input model with a 3-D VESSEL component, the vessel and downcomer inventories during 3-IB-3 were generally well predicted, but the core heatup was underpredicted compared to data. An equivalent calculation with an all 1-D input model ran about twice as fast as our basecase analysis using a 3-D VESSEL in the input model, but the results of the two calculations diverged significantly for many parameters of interest, with the 3-D VESSEL model results in better agreement with data. 22 refs., 100 figs

  2. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the FRAP code (Fuel Rod Analysis Program) and applied to a PWR fuel rod undergoing a LOCA. The method of uncertainty analysis is the Response Surface Method (RSM). (author)

  3. A friend man-machine interface for thermo-hydraulic simulation codes of nuclear installations

    International Nuclear Information System (INIS)

    Araujo Filho, F. de; Belchior Junior, A.; Barroso, A.C.O.; Gebrim, A.

    1994-01-01

    This work presents the development of a Man-Machine Interface to the TRAC-PF1 code, a computer program to perform best estimate analysis of transients and accidents at nuclear power plants. The results were considered satisfactory and a considerable productivity gain was achieved in the activity of preparing and analyzing simulations. (author)

  4. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  5. MELPROG/TRAC: update and applications

    International Nuclear Information System (INIS)

    Henninger, R.J.; Kelly, J.E.

    1986-01-01

    The first complete, coupled, and mechanistic analysis of a core meltdown sequence has been made with MELPROG-PWR/MOD1 and MELPROG/TRAC. The sequence analyzed was a station blackout accident for the Surry plant. Through vessel failure, all important aspects of the meltdown sequence were calculated. This version of MELPROG permits a full two-dimensional treatment of the in-vessel phenomena. Natural circulation can thus be modeled. Comparison to one-dimensional MELPROG and MARCH calculations shows that natural circulation reduces the rate of core heating, but increases the rate of heating of upper plenum structures and primary piping. This increased heating can inhibit fission product deposition and may lead to an early failure of the primary system. Because of uncertainty, sensitivity studies were performed to assess the relative importance of modeling assumptions. Changes in the modeling of the initial fuel rod melting and relocation were found to vary the hydrogen source by a factor of 2 and alter the timing of key events. These results imply that accurate and mechanistic modeling is important for severe accident sequence analysis

  6. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

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

    1996-01-01

    Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

  7. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  8. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  9. Comparison of a TRAC calculation to the data from LSTF run SB-CL-05

    International Nuclear Information System (INIS)

    Motley, F.; Schultz, R.

    1986-01-01

    Run SB-CL-05 is a 5% break in the side of the cold leg. The test results show that the core was uncovered briefly and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam generator tubes was observed. When the loop seal cleared, the core refilled and the rods cooled. The TRAC results are in reasonable agreement with the test data, meaning that TRAC correctly predicted the major trends and phenomena. TRAC predicted the core uncovery, the resulting rod heatup, and the liquid holdup on the upflow side of the steam generator tubes correctly. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data, but TRAC predicted its occurrence 20 s late. The experimental and TRAC analysis results of run SB-CL-05 are similar to those for Semiscale Run S-UT-8. In both runs there was core uncovery, rod overheating, and steam generator liquid holdup. These results confirm scaling of these phenomena from Semiscale (1/1650) to LSTF (1/48)

  10. A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics

    International Nuclear Information System (INIS)

    Lopez, Arsenio Procopio; Morales Sandoval, Jaime B.

    2005-01-01

    The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible

  11. ENSDF analysis codes. IBM version, August 1982

    International Nuclear Information System (INIS)

    Lorenz, A.

    1982-01-01

    The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)

  12. ENSDF analysis codes: IBM version. August 1982

    International Nuclear Information System (INIS)

    Lorenz, A.

    1982-09-01

    The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)

  13. TRAC analyses for CCTF and SCTF tests and UPTF design/operation

    International Nuclear Information System (INIS)

    Williams, K.A.

    1983-01-01

    The 2D/3D Program is a multinational (Germany, Japan, and the United States) experimental and analytical nuclear reactor safety research program. The Los Alamos analysis effort is functioning as a vital part of the 2D/3D program. The CCTF and SCTF analyses have demonstrated that TRAC-PF1 can correctly predict multidimensional, nonequilibrium behavior in large-scale facilities prototypical of actual PWR's. Through these and future TRAC analyses the experimental findings can be related from facility to facility, and the results of this research program can be directly related to licensing concerns affecting actual PWR's

  14. Hazardous chemical tracking system (HAZ-TRAC)

    International Nuclear Information System (INIS)

    Bramlette, J.D.; Ewart, S.M.; Jones, C.E.

    1990-07-01

    Westinghouse Idaho Nuclear Company, Inc. (WINCO) developed and implemented a computerized hazardous chemical tracking system, referred to as Haz-Trac, for use at the Idaho Chemical Processing Plant (ICPP). Haz-Trac is designed to provide a means to improve the accuracy and reliability of chemical information, which enhances the overall quality and safety of ICPP operations. The system tracks all chemicals and chemical components from the time they enter the ICPP until the chemical changes form, is used, or becomes a waste. The system runs on a Hewlett-Packard (HP) 3000 Series 70 computer. The system is written in COBOL and uses VIEW/3000, TurboIMAGE/DBMS 3000, OMNIDEX, and SPEEDWARE. The HP 3000 may be accessed throughout the ICPP, and from remote locations, using data communication lines. Haz-Trac went into production in October, 1989. Currently, over 1910 chemicals and chemical components are tracked on the system. More than 2500 personnel hours were saved during the first six months of operation. Cost savings have been realized by reducing the time needed to collect and compile reporting information, identifying and disposing of unneeded chemicals, and eliminating duplicate inventories. Haz-Trac maintains information required by the Superfund Amendment Reauthorization Act (SARA), the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) and the Occupational Safety and Health Administration (OSHA)

  15. Hazardous chemical tracking system (HAZ-TRAC)

    Energy Technology Data Exchange (ETDEWEB)

    Bramlette, J D; Ewart, S M; Jones, C E

    1990-07-01

    Westinghouse Idaho Nuclear Company, Inc. (WINCO) developed and implemented a computerized hazardous chemical tracking system, referred to as Haz-Trac, for use at the Idaho Chemical Processing Plant (ICPP). Haz-Trac is designed to provide a means to improve the accuracy and reliability of chemical information, which enhances the overall quality and safety of ICPP operations. The system tracks all chemicals and chemical components from the time they enter the ICPP until the chemical changes form, is used, or becomes a waste. The system runs on a Hewlett-Packard (HP) 3000 Series 70 computer. The system is written in COBOL and uses VIEW/3000, TurboIMAGE/DBMS 3000, OMNIDEX, and SPEEDWARE. The HP 3000 may be accessed throughout the ICPP, and from remote locations, using data communication lines. Haz-Trac went into production in October, 1989. Currently, over 1910 chemicals and chemical components are tracked on the system. More than 2500 personnel hours were saved during the first six months of operation. Cost savings have been realized by reducing the time needed to collect and compile reporting information, identifying and disposing of unneeded chemicals, and eliminating duplicate inventories. Haz-Trac maintains information required by the Superfund Amendment Reauthorization Act (SARA), the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) and the Occupational Safety and Health Administration (OSHA).

  16. R-matrix analysis code (RAC)

    International Nuclear Information System (INIS)

    Chen Zhenpeng; Qi Huiquan

    1990-01-01

    A comprehensive R-matrix analysis code has been developed. It is based on the multichannel and multilevel R-matrix theory and runs in VAX computer with FORTRAN-77. With this code many kinds of experimental data for one nuclear system can be fitted simultaneously. The comparisions between code RAC and code EDA of LANL are made. The data show both codes produced the same calculation results when one set of R-matrix parameters was used. The differential cross section of 10 B (n, α) 7 Li for E n = 0.4 MeV and the polarization of 16 O (n,n) 16 O for E n = 2.56 MeV are presented

  17. TOOKUIL: A case study in user interface development for safety code application

    International Nuclear Information System (INIS)

    Gray, D.L.; Harkins, C.K.; Hoole, J.G.

    1997-01-01

    Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today's safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL

  18. TOOKUIL: A case study in user interface development for safety code application

    International Nuclear Information System (INIS)

    Gray, D.L.; Harkins, C.K.; Hoole, J.G.; Peebles, R.C.; Smith, R.J.

    1996-11-01

    Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today's safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL

  19. TOOKUIL: A case study in user interface development for safety code application

    Energy Technology Data Exchange (ETDEWEB)

    Gray, D.L.; Harkins, C.K.; Hoole, J.G. [and others

    1997-07-01

    Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today`s safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL.

  20. First intermediate break test 6IB1 data comparison with a TRAC-BD1/MOD1 blind calculation

    International Nuclear Information System (INIS)

    Wheatley, P.D.

    1985-04-01

    TRAC-BD1/MOD1 has been used to calculate the behavior in the FIST (Full Integral Test Facility) facility during an intermediate break in one of the recirculation loops. Results of the calculation are compared with the data from the experiment, and the analysis is discussed in this report. The calculation was blind with only the initial and boundary conditions available prior to performance of the calculation. The calculation has been previously documented without reference to the experimental data (i.e., prior to release of the data). This report extends the prior report by discussing the analysis of the data to code comparisons. This work was performed as part of the Nuclear Regulatory Commission's support to the FIST program which is being provided at the Idaho National Engineering Laboratory

  1. Centrifugal Compressor Aeroelastic Analysis Code

    Science.gov (United States)

    Keith, Theo G., Jr.; Srivastava, Rakesh

    2002-01-01

    Centrifugal compressors are very widely used in the turbomachine industry where low mass flow rates are required. Gas turbine engines for tanks, rotorcraft and small jets rely extensively on centrifugal compressors for rugged and compact design. These compressors experience problems related with unsteadiness of flowfields, such as stall flutter, separation at the trailing edge over diffuser guide vanes, tip vortex unsteadiness, etc., leading to rotating stall and surge. Considerable interest exists in small gas turbine engine manufacturers to understand and eventually eliminate the problems related to centrifugal compressors. The geometric complexity of centrifugal compressor blades and the twisting of the blade passages makes the linear methods inapplicable. Advanced computational fluid dynamics (CFD) methods are needed for accurate unsteady aerodynamic and aeroelastic analysis of centrifugal compressors. Most of the current day industrial turbomachines and small aircraft engines are designed with a centrifugal compressor. With such a large customer base and NASA Glenn Research Center being, the lead center for turbomachines, it is important that adequate emphasis be placed on this area as well. Currently, this activity is not supported under any project at NASA Glenn.

  2. Laguna Verde simulator: A new TRAC-RT based application

    International Nuclear Information System (INIS)

    Munoz Cases, J.J.; Tanarro Onrubia, A.

    2006-01-01

    In a partnership with GSE Systems, TECNATOM is developing a full scope training simulator for Laguna Verde Unit 2 (LV2). The simulator design is based upon the current 'state-of-the art technology' regarding the simulation platform, instructor station, visualization tools, advanced thermalhydraulics and neutronics models, I/O systems and automated model building technology. When completed, LV2 simulator will achieve a remarkable level of modeling fidelity by using TECNATOM's TRAC-RT advanced thermalhydraulic code for the reactor coolant and main steam systems, and NEMO neutronic model for the reactor core calculations. These models have been utilized up to date for the development or upgrading of nine NPP simulators in Spain and abroad, with more than 8000 hours of training sessions, and have developed an excellent reputation for its robustness and high fidelity. (author)

  3. NORTICA - a new code for cyclotron analysis

    International Nuclear Information System (INIS)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-01-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state

  4. Assessment of TRAC-BD1/MOD1 using FIST data

    International Nuclear Information System (INIS)

    Jo, J.H.; Connell, H.R.

    1985-01-01

    This report is concerned with the assessment of the TRAC-BD1/MOD1 Code, developed at Idaho National Engineering Laboratory. The assessment was conducted using data from the FIST (Full Integral Simulation Test) facility, which is a BWR safety test facility which was built to investigate small break LOCA and operational transients in BWR's and to complement earlier large break LOCA test results from TLTA (Two-Loop Test Apparatus). 21 figs

  5. CATHENA 4. A thermalhydraulics network analysis code

    International Nuclear Information System (INIS)

    Aydemir, N.U.; Hanna, B.N.

    2009-01-01

    Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA) is a one-dimensional, non-equilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. The objective of the present paper is to describe the design, application and future development plans for the CATHENA 4 thermalhydraulics network analysis code, which is a modernized version of the present frozen CATHENA 3 code. The new code is designed in modular form, using the Fortran 95 (F95) programming language. The semi-implicit numerical integration scheme of CATHENA 3 is re-written to implement a fully-implicit methodology using Newton's iterative solution scheme suitable for nonlinear equations. The closure relations, as a first step, have been converted from the existing CATHENA 3 implementation to F95 but modularized to achieve ease of maintenance. The paper presents the field equations, followed by a description of the Newton's scheme used. The finite-difference form of the field equations is given, followed by a discussion of convergence criteria. Two applications of CATHENA 4 are presented to demonstrate the temporal and spatial convergence of the new code for problems with known solutions or available experimental data. (author)

  6. POST: a postprocessor computer code for producing three-dimensional movies of two-phase flow in a reactor vessel

    International Nuclear Information System (INIS)

    Taggart, K.A.; Liles, D.R.

    1977-08-01

    The development of the TRAC computer code for analysis of LOCAs in light-water reactors involves the use of a three-dimensional (r-theta-z), two-fluid hydrodynamics model to describe the two-phase flow of steam and water through the reactor vessel. One of the major problems involved in interpreting results from this code is the presentation of three-dimensional flow patterns. The purpose of the report is to present a partial solution to this data display problem. A first version of a code which produces three-dimensional movies of flow in the reactor vessel has been written and debugged. This code (POST) is used as a postprocessor in conjunction with a stand alone three-dimensional two-phase hydrodynamics code (CYLTF) which is a test bed for the three-dimensional algorithms to be used in TRAC

  7. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  8. Static Code Analysis with Gitlab-CI

    CERN Document Server

    Datko, Szymon Tomasz

    2016-01-01

    Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...

  9. Turbo Pascal Computer Code for PIXE Analysis

    International Nuclear Information System (INIS)

    Darsono

    2002-01-01

    To optimal utilization of the 150 kV ion accelerator facilities and to govern the analysis technique using ion accelerator, the research and development of low energy PIXE technology has been done. The R and D for hardware of the low energy PIXE installation in P3TM have been carried on since year 2000. To support the R and D of PIXE accelerator facilities in harmonize with the R and D of the PIXE hardware, the development of PIXE software for analysis is also needed. The development of database of PIXE software for analysis using turbo Pascal computer code is reported in this paper. This computer code computes the ionization cross-section, the fluorescence yield, and the stopping power of elements also it computes the coefficient attenuation of X- rays energy. The computer code is named PIXEDASIS and it is part of big computer code planed for PIXE analysis that will be constructed in the near future. PIXEDASIS is designed to be communicative with the user. It has the input from the keyboard. The output shows in the PC monitor, which also can be printed. The performance test of the PIXEDASIS shows that it can be operated well and it can provide data agreement with data form other literatures. (author)

  10. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the Fuel Rod Analysis Program (FRAP) code and has been applied to a pressurized water reactor (PWR) fuel rod undergoing a loss-of-coolant accident (LOCA). The method of uncertainty analysis is the response surface method. The automated version significantly reduced the time required to complete the analysis and, at the same time, greatly increased the problem scope. Results of the analysis showed a significant difference in the total and relative contributions to the uncertainty of the response parameters between steady state and transient conditions

  11. Current algorithms used in reactor safety codes and the impact of future computer development on these algorithms

    International Nuclear Information System (INIS)

    Mahaffy, J.H.; Liles, D.R.; Woodruff, S.B.

    1985-01-01

    Computational methods and solution procedures used in the US Nuclear Regulatory Commission's reactor safety systems codes, Transient Reactor Analysis Code (TRAC) and Reactor Leak and Power Safety Excursion Code (RELAP), are reviewed. Methods used in TRAC-PF1/MOD1, including the stability-enhancing two-step (SETS) technique, which permits fast computations by allowing time steps larger than the material Courant stability limit, are described in detail, and the differences from RELAP5/MOD2 are noted. Developments in computing, including parallel and vector processing, and their applicability to nuclear reactor safety codes are described. These developments, coupled with appropriate numerical methods, make detailed faster-than-real-time reactor safety analysis a realistic near-term possibility

  12. NESTLE: A nodal kinetics code

    International Nuclear Information System (INIS)

    Al-Chalabi, R.M.; Turinsky, P.J.; Faure, F.-X.; Sarsour, H.N.; Engrand, P.R.

    1993-01-01

    The NESTLE nodal kinetics code has been developed for utilization as a stand-alone code for steady-state and transient reactor neutronic analysis and for incorporation into system transient codes, such as TRAC and RELAP. The latter is desirable to increase the simulation fidelity over that obtained from currently employed zero- and one-dimensional neutronic models and now feasible due to advances in computer performance and efficiency of nodal methods. As a stand-alone code, requirements are that it operate on a range of computing platforms from memory-limited personal computers (PCs) to supercomputers with vector processors. This paper summarizes the features of NESTLE that reflect the utilization and requirements just noted

  13. Benchmark calculation of subchannel analysis codes

    International Nuclear Information System (INIS)

    1996-02-01

    In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

  14. Sandia National Laboratories analysis code data base

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, C.W.

    1994-11-01

    Sandia National Laboratories, mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The Laboratories` strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia`s technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code ``ownership`` and release status, and references describing the physical models and numerical implementation.

  15. Sandia National Laboratories analysis code data base

    Science.gov (United States)

    Peterson, C. W.

    1994-11-01

    Sandia National Laboratories' mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The laboratories' strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia's technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems, and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code 'ownership' and release status, and references describing the physical models and numerical implementation.

  16. Development of chemical equilibrium analysis code 'CHEEQ'

    International Nuclear Information System (INIS)

    Nagai, Shuichiro

    2006-08-01

    'CHEEQ' code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. Characteristics of 'CHEEQ' code are as follows. All the chemical equilibrium equations were described by the formation reactions from the mono-atomic gases in order to simplify the code structure and input preparation. Chemical equilibrium conditions, Σν i μ i =0 for the gaseous compounds and precipitated condensed phase compounds and Σν i μ i > 0 for the non-precipitated condensed phase compounds, were applied. Where, ν i and μ i are stoichiometric coefficient and chemical potential of component i. Virtual solid model was introduced to perform the calculation of constant partial pressure condition. 'CHEEQ' was consisted of following 3 parts, (1) analysis code, zc132. f. (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. 'CHEEQ' code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200). Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instructions to prepare the input data and to evaluate the calculation results. (author)

  17. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

    Parsa, Z.

    1988-01-01

    We present a comparison between results obtained from standard accelerator physics codes used for the design and analysis of synchrotrons and storage rings, with programs SYNCH, MAD, HARMON, PATRICIA, PATPET, BETA, DIMAD, MARYLIE and RACE-TRACK. In our analysis we have considered 5 (various size) lattices with large and small angles including AGS Booster (10/degree/ bend), RHIC (2.24/degree/), SXLS, XLS (XUV ring with 45/degree/ bend) and X-RAY rings. The differences in the integration methods used and the treatment of the fringe fields in these codes could lead to different results. The inclusion of nonlinear (e.g., dipole) terms may be necessary in these calculations specially for a small ring. 12 refs., 6 figs., 10 tabs

  18. Electromagnetic field and mechanical stress analysis code

    International Nuclear Information System (INIS)

    1978-01-01

    Analysis TEXMAGST is a two stage linear finite element code for the analysis of static magnetic fields in three dimensional structures and associated mechanical stresses produced by the anti J x anti B forces within these structures. The electromagnetic problem is solved in terms of magnetic vector potential A for a given current density anti J as curl 1/μ curl anti A = anti J considering the magnetic permeability as constant. The Coulombian gauge (div anti A = o) was chosen and was implemented through the use of Lagrange multipliers. The second stage of the problem - the calculation of mechanical stresses in the same three dimensional structure is solved by using the same code with few modifications - through a restart card. Body forces anti J x anti B within each element are calculated from the solution of the first stage run and represent the input to the second stage run which will give the solution for the stress problem

  19. Fire-accident analysis code (FIRAC) verification

    International Nuclear Information System (INIS)

    Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.

    1986-01-01

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

  20. Web interface for plasma analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Emoto, M. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)], E-mail: emo@nifs.ac.jp; Murakami, S. [Kyoto University, Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan); Yoshida, M.; Funaba, H.; Nagayama, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)

    2008-04-15

    There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach.

  1. Web interface for plasma analysis codes

    International Nuclear Information System (INIS)

    Emoto, M.; Murakami, S.; Yoshida, M.; Funaba, H.; Nagayama, Y.

    2008-01-01

    There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach

  2. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  3. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  4. Development of disruption thermal analysis code DREAM

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro

    1989-07-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).

  5. Development of disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.

    1989-01-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)

  6. Investigation of Two-Phase Flow Regime Maps for Development of Thermal-Hydraulic Analysis Codes

    International Nuclear Information System (INIS)

    Kim, Kyung Doo; Kim, Byoung Jae; Lee, Seong Wook

    2010-04-01

    This reports is a literature survey on models and correlations for determining flow pattern that are used to simulate thermal-hydraulics in nuclear reactors. Determination of flow patterns are a basis for obtaining physical values of wall/interfacial friction, wall/interfacial heat transfer, and droplet entrainment/de-entrainment. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed to find models and correlations

  7. CONTEMPT-DG containment analysis code

    International Nuclear Information System (INIS)

    Deem, R.E.; Rousseau, K.

    1982-01-01

    The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given

  8. Extending CANTUP code analysis to probabilistic evaluations

    International Nuclear Information System (INIS)

    Florea, S.

    2001-01-01

    The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented

  9. Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

    International Nuclear Information System (INIS)

    Schwegler, E.C.; Maudlin, P.J.

    1983-01-01

    A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

  10. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1988-01-01

    This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared

  11. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  12. Prediction of reflood behavior for tests with differing axial power shapes using WCOBRA/TRAC

    International Nuclear Information System (INIS)

    Bajorek, S.M.; Hochreiter, L.E.

    1991-01-01

    The rector core power shape can vary over the fuel cycle due to load follow, control rod movement, burnup effects and Xenon transients. a best estimate thermal-hydraulic code must be able to accurately predict the reflooding behavior for different axial power shapes in order to find the power shapes effects on the loss-of-coolant peak cladding temperature. Several different reflood heat transfer experiments have been performed at the same or similar PWR reflood conditions with different axial power shapes. These experiments have different rod diameters, were full length, 3.65 m (12 feet) in height, and had simple egg crate grids. The WCOBRA/TRAC code has been used to model several different tests from these three experiments to examine the code's capability to predict the reflood transient for different power shapes, with a consistent model and noding scheme. This paper describes these different experiments, their power shapes, and the test conditions. The WCOBRA/TRAC code is described as well as the noding scheme, and the calculated results will be compared in detail with the test data rod temperatures. An overall assessment of the code's predictions of these experiments is presented

  13. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  14. Current lead thermal analysis code 'CURRENT'

    International Nuclear Information System (INIS)

    Yamaguchi, Masahito; Tada, Eisuke; Shimamoto, Susumu; Hata, Kenichiro.

    1985-08-01

    Large gas-cooled current lead with the capacity more than 30 kA and 22 kV is required for superconducting toroidal and poloidal coils for fusion application. The current lead is used to carry electrical current from the power supply system at room temperature to the superconducting coil at 4 K. Accordingly, the thermal performance of the current lead is significantly important to determine the heat load requirements of the coil system at 4 K. Japan Atomic Energy Research Institute (JAERI) has being developed the large gas-cooled current leads with the optimum condition in which the heat load is around 1 W per 1 kA at 4 K. In order to design the current lead with the optimum thermal performances, JAERI developed thermal analysis code named as ''CURRENT'' which can theoretically calculate the optimum geometric shape and cooling conditions of the current lead. The basic equations and the instruction manual of the analysis code are described in this report. (author)

  15. Assessment of selected TRAC and RELAP5 calculations for Oconee-1 pressurized thermal shock study

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Pu, J.; Saha, P.; Jo, J.

    1984-11-01

    Several Oconee-1 overcooling transients that were computed by LANL and INEL using the latest versions of TRAC-PF1 and RELAPS/MOD1.5 codes have been reviewed by BNL. Three of these transients were selected for detailed review as they either had the potential of challenging the integrity of the pressure vessel or highlighted the effect of code differences. These are: (1) Main Steam Line Break (MSLB); (2) All Turbine Bypass Valves Stuck Open; and (3) 2-Inch Small Break LOCA. Both codes were reasonably successful in modeling these transients. However, there were differences in the code results even though the specified scenarios were exactly the same for two transients (MSLB and Small Break LOCA). This report compares the code results and explains the possible reasons for these differences. Recommendations have been made regarding which result seems more reasonable for a specific transient

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

    International Nuclear Information System (INIS)

    Queral, C.; Mulas, J.; Collazo, I.; Concejal, A.; Burbano, N.; Gallego, I.; Lopez Lechas, A.

    2002-01-01

    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)

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

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

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

  18. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    International Nuclear Information System (INIS)

    Aksan, S.N.; Stierli, F.; Analytis, G.T.

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended

  19. Error-correction coding and decoding bounds, codes, decoders, analysis and applications

    CERN Document Server

    Tomlinson, Martin; Ambroze, Marcel A; Ahmed, Mohammed; Jibril, Mubarak

    2017-01-01

    This book discusses both the theory and practical applications of self-correcting data, commonly known as error-correcting codes. The applications included demonstrate the importance of these codes in a wide range of everyday technologies, from smartphones to secure communications and transactions. Written in a readily understandable style, the book presents the authors’ twenty-five years of research organized into five parts: Part I is concerned with the theoretical performance attainable by using error correcting codes to achieve communications efficiency in digital communications systems. Part II explores the construction of error-correcting codes and explains the different families of codes and how they are designed. Techniques are described for producing the very best codes. Part III addresses the analysis of low-density parity-check (LDPC) codes, primarily to calculate their stopping sets and low-weight codeword spectrum which determines the performance of these codes. Part IV deals with decoders desi...

  20. 76 FR 64931 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-10-19

    ...-0046] Building Energy Codes Cost Analysis AGENCY: Office of Energy Efficiency and Renewable Energy... reopening of the time period for submitting comments on the request for information on Building Energy Codes... the request for information on Building Energy Code Cost Analysis and provide docket number EERE-2011...

  1. TRAC Innovation Report (FY15 to FY18)

    Science.gov (United States)

    2017-11-01

    Operations and Reports, 1215 Jefferson Davis Highway, Suite 1204, Arlington, VA 22202-4302, and to the Office of Management and Budget, Paperwork Reduction...inside and outside of TRAC. Rapid, successful adoption of ideas requires leadership and support, but innovation is fundamentally a bottom up phenomena...the organizational environment. TRAC provides a knowledge management system and communities of interest and practice to foster teamwork

  2. Financial Health of the Higher Education Sector: Financial Results and TRAC Outcomes 2013-14. Issues Paper 2015/07

    Science.gov (United States)

    Higher Education Funding Council for England, 2015

    2015-01-01

    This report provides an overview of the financial health of the Higher Education Funding Council for England (HEFCE)-funded higher education sector in England. The analysis covers financial results for the academic year 2013-14, as submitted to HEFCE in December 2014, as well as the outcomes from the sector's Transparent Approach to Costing (TRAC)…

  3. Preliminary Study of 1D Thermal-Hydraulic System Analysis Code Using the Higher-Order Numerical Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Woong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The existing nuclear system analysis codes such as RELAP5, TRAC, MARS and SPACE use the first-order numerical scheme in both space and time discretization. However, the first-order scheme is highly diffusive and less accurate due to the first order of truncation error. So, the numerical diffusion problem which makes the gradients to be smooth in the regions where the gradients should be steep can occur during the analysis, which often predicts less conservatively than the reality. Therefore, the first-order scheme is not always useful in many applications such as boron solute transport. RELAP7 which is an advanced nuclear reactor system safety analysis code using the second-order numerical scheme in temporal and spatial discretization is being developed by INL (Idaho National Laboratory) since 2011. Therefore, for better predictive performance of the safety of nuclear reactor systems, more accurate nuclear reactor system analysis code is needed for Korea too to follow the global trend of nuclear safety analysis. Thus, this study will evaluate the feasibility of applying the higher-order numerical scheme to the next generation nuclear system analysis code to provide the basis for the better nuclear system analysis code development. The accuracy is enhanced in the spatial second-order scheme and the numerical diffusion problem is alleviated while indicates significantly lower maximum Courant limit and the numerical dispersion issue which produces spurious oscillation and non-physical results in the higher-order scheme. If the spatial scheme is the first order scheme then the temporal second-order scheme provides almost the same result with the temporal firstorder scheme. However, when the temporal second order scheme and the spatial second-order scheme are applied together, the numerical dispersion can occur more severely. For the more in-depth study, the verification and validation of the NTS code built in MATLAB will be conducted further and expanded to handle two

  4. Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

    Science.gov (United States)

    Harrington, James William

    Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

  5. Best estimate LB LOCA approach based on advanced thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.

    2004-01-01

    Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)

  6. Evaluation of accuracy in the ExacTrac 6D image induced radiotherapy using CBCT

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Chun; Kim, Hyo Jung; Kim, Jong Deok; Ji, Dong Hwa; Song, Ju Young [Dept. of Radiation Oncology, Chonnam National University School, Gwangju (Korea, Republic of)

    2016-12-15

    To verify the accuracy of the image guided radiotherapy using ExacTrac 6D couch, the error values in six directions are randomly assigned and corrected and then the corrected values were compared with CBCT image to check the accurateness of ExacTrac. The therapy coordination values in the Rando head Phantom were moved in the directions of X, Y and Z as the translation group and they were moved in the directions of pitch, roll and yaw as the rotation group. The corrected values were moved in 6 directions with the combined and mutual reactions. The Z corrected value ranges from 1mm to 23mm. In the analysis of errors between CBCT image of the phantom which is corrected with therapy coordinate and 3D/3D matching error value, the rotation group showed higher error value than the translation group. In the distribution of dose for the error value of the therapy coordinate corrected with CBCT, the restricted value of dosage for the normal organs in two groups meet the prescription dose. In terms of PHI and PCI values which are the dose homogeneity of the cancerous tissue, the rotation group showed a little higher in the low dose distribution range. This study is designed to verify the accuracy of ExacTrac 6D couch using CBCT. It showed that in terms of the error value in the simple movement, it showed the comparatively accurate correction capability but in the movement when the angle is put in the couch, it showed the inaccurate correction values. So, if the body of the patient is likely to have a lot of changes in the direction of rotation or there is a lot of errors in the pitch, roll and yaw in ExacTrac correction, it is better to conduct the CBCT guided image to correct the therapy coordinate in order to minimize any side effects.

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

    Science.gov (United States)

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

    2018-01-01

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

  8. Application of coupled codes for safety analysis and licensing issues

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    2006-01-01

    An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

  9. Analytical validation of the CACECO containment analysis code

    International Nuclear Information System (INIS)

    Peak, R.D.

    1979-08-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code

  10. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  12. Uranium Isotopic Analysis with the FRAM Isotopic Analysis Code

    International Nuclear Information System (INIS)

    Vo, D.T.; Sampson, T.E.

    1999-01-01

    FRAM is the acronym for Fixed-Energy Response-Function Analysis with Multiple efficiency. This software was developed at Los Alamos National Laboratory originally for plutonium isotopic analysis. Later, it was adapted for uranium isotopic analysis in addition to plutonium. It is a code based on a self-calibration using several gamma-ray peaks for determining the isotopic ratios. The versatile-parameter database structure governs all facets of the data analysis. User editing of the parameter sets allows great flexibility in handling data with different isotopic distributions, interfering isotopes, and different acquisition parameters such as energy calibration and detector type

  13. Codeword Structure Analysis for LDPC Convolutional Codes

    Directory of Open Access Journals (Sweden)

    Hua Zhou

    2015-12-01

    Full Text Available The codewords of a low-density parity-check (LDPC convolutional code (LDPC-CC are characterised into structured and non-structured. The number of the structured codewords is dominated by the size of the polynomial syndrome former matrix H T ( D , while the number of the non-structured ones depends on the particular monomials or polynomials in H T ( D . By evaluating the relationship of the codewords between the mother code and its super codes, the low weight non-structured codewords in the super codes can be eliminated by appropriately choosing the monomials or polynomials in H T ( D , resulting in improved distance spectrum of the mother code.

  14. TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Sekhri, A.; Baumann, P. [KernkraftwerkLeibstadt AG, 5325 Leibstadt (Switzerland); Wicaksono, D. [Swiss Federal Inst. of Technology Zurich ETH, 8092 Zurich (Switzerland); Miro, R.; Barrachina, T.; Verdu, G. [Inst. for Industrial, Radiophysical and Environmental Safety ISIRYM, Universitat Politecnica de Valencia UPV, Cami de Vera s/n, 46021 Valencia (Spain)

    2012-07-01

    In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

  15. TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

    International Nuclear Information System (INIS)

    Sekhri, A.; Baumann, P.; Wicaksono, D.; Miro, R.; Barrachina, T.; Verdu, G.

    2012-01-01

    In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

  16. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

  17. Parallelization of Subchannel Analysis Code MATRA

    International Nuclear Information System (INIS)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk

    2014-01-01

    A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems

  18. Module type plant system dynamics analysis code (MSG-COPD). Code manual

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2002-11-01

    MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

  19. OPR1000 RCP Flow Coastdown Analysis using SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

  20. Fast neutron analysis code SAD1

    International Nuclear Information System (INIS)

    Jung, M.; Ott, C.

    1985-01-01

    A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum

  1. Summary of transient analysis

    International Nuclear Information System (INIS)

    Saha, P.

    1984-01-01

    This chapter reviews the papers on the pressurized water reactor (PWR) and boiling water reactor (BWR) transient analyses given at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Most of the papers were based on the systems calculations performed using the TRAC-PWR, RELAP5 and RETRAN codes. The status of the nuclear industry in the code applications area is discussed. It is concluded that even though comprehensive computer codes are available for plant transient analysis, there is still a need to exercise engineering judgment, simpler tools and even hand calculations to supplement these codes

  2. Assessment of MSIV full closure for Santa Maria de Garona Nuclear Power Plant using TRAC-BF1 (G1J1)

    International Nuclear Information System (INIS)

    Crespo, J.L.

    1993-06-01

    This document presents a spurious Main Steam Isolation Value (MSIV) closure analysis for Santa Maria de Garorta Nuclear Power Plan describing the problems found when comparing calculated and real data. The plant is a General Electric Boiling Water Reactor 3, containment type Mark 1. It is operated by NUCLENOR, S.A. and was connected to the grid in 1971. The analysis has been performed by the Apphed Physics Department from the University of Cantabria and the Analysis and Operation Section from NUCLENOR, S.A. as a part of an agreement for developing an engineering simulator of operational transients and accidents for Santa Maria de Gamma Power Plant. The analysis was performed using the frozen version of TRAC-BFI (GlJl) code and is the second of two NUCLENOR contributions to the International Code Applications and Assessment Program (ICAP). The code was run in a Cyber 932 with operating system NOS/VE, property of NUCLENOR, S.A.. A programming effort was carried out in order to provide suitable graphics from the output file

  3. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  4. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  5. ZERBERUS - the code for reliability analysis of crack containing structures

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.

    1992-04-01

    Brief description of the First- and Second Order Reliability Methods, being the theoretical background of the code, is given. The code structure is described in detail, with special emphasis to the new application fields. The numerical example investigates failure probability of steam generator tubing affected by stress corrosion cracking. The changes necessary to accommodate this analysis within the ZERBERUS code are explained. Analysis results are compared to different Monte Carlo techniques. (orig./HP) [de

  6. Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

  7. Content Analysis Coding Schemes for Online Asynchronous Discussion

    Science.gov (United States)

    Weltzer-Ward, Lisa

    2011-01-01

    Purpose: Researchers commonly utilize coding-based analysis of classroom asynchronous discussion contributions as part of studies of online learning and instruction. However, this analysis is inconsistent from study to study with over 50 coding schemes and procedures applied in the last eight years. The aim of this article is to provide a basis…

  8. A Semantic Analysis Method for Scientific and Engineering Code

    Science.gov (United States)

    Stewart, Mark E. M.

    1998-01-01

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

  9. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

  10. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  11. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  12. Evaluation of the DRAGON code for VHTR design analysis

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-01

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR

  13. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  14. TracWorks - global fuel assembly data management

    International Nuclear Information System (INIS)

    Cooney, B.F.

    1997-01-01

    The TracWorks Data Management System is a workstation-based software product that provides a utility with a single, broadly available, regularly updated source for virtually every data item available for a fuel assembly or core component. TracWorks is designed to collect, maintain and provide information about assembly and component locations and movements during the refuelling process and operation, assembly burnup and isotopic inventory (both in-core and out-of-core), pin burnup and isotopics for pins that have been removed from their original assemblies, assembly and component inspection results (including video) and manufacturing data provided by the fabrication plant. (UK)

  15. TracWorksTM-global fuel assembly data management

    International Nuclear Information System (INIS)

    Cooney, B.F.

    1997-01-01

    The TracWorks TM Data Management System is a workstation-based software product that provides a utility with a single, broadly available, regularly updated source for virtually every data item available for a fuel assembly or core component. TracWorks is designed to collect, maintain and provide information about assembly and component locations and movements during the refueling process and operation, assembly burnup and isotopic inventory (both in-core and out-of-core), pin burnup and isotopics for pins that have been removed from their original assemblies, assembly and component inspection results (including video) and manufacturing data provided by the fabrication plant

  16. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1983-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  17. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1982-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  18. Posttest TRAC-PD2/MOD1 predictions for FLECHT SEASET test 31504

    International Nuclear Information System (INIS)

    Booker, C.P.

    1982-01-01

    TRAC-PD2/MOD1 is a publicly released version of TRAC that is used primarily to analyze large-break loss-of-coolant accidents in pressurized-water reactors (PWRs). TRAC-PD2 can calculate, among other things, reflood phenomena. TRAC posttest predictions are compared with test 31504 reflood data from the Full-Length Emergency Core Heat Transfer (FLECHT) System Effects and Separate Effects Tests (SEASET) facility. A false top-down quench is predicted near the top of the core and the subcooling is underpredicted at the bottom of the core. However, the overall TRAC predictions are good, especially near the center of the core

  19. Comparison of Interfacial and Wall Friction Models in Thermal-Hydraulic System Analysis Codes (Rev1.0)

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Kim, Soo Hyung; Kim, Byung Jae; Chung, Bub Dong; Kim, Hee Cheol

    2010-04-01

    This reports is a literature survey on models and correlations for interfacial and wall friction models that are used to simulate thermal-hydraulics in nuclear reactors. The interfacial and wall frictions are needed to solve the momentum equations of gas, continuous liquid and droplet. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed. This report is a revised version of the previous technical report(KAERI/TR-3437/2007)

  20. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

    Han, S.H.; Yoo, K.J.; Kim, T.W.

    1987-01-01

    The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

  1. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  2. A Comparative Study on Seismic Analysis of Bangladesh National Building Code (BNBC) with Other Building Codes

    Science.gov (United States)

    Bari, Md. S.; Das, T.

    2013-09-01

    Tectonic framework of Bangladesh and adjoining areas indicate that Bangladesh lies well within an active seismic zone. The after effect of earthquake is more severe in an underdeveloped and a densely populated country like ours than any other developed countries. Bangladesh National Building Code (BNBC) was first established in 1993 to provide guidelines for design and construction of new structure subject to earthquake ground motions in order to minimize the risk to life for all structures. A revision of BNBC 1993 is undergoing to make this up to date with other international building codes. This paper aims at the comparison of various provisions of seismic analysis as given in building codes of different countries. This comparison will give an idea regarding where our country stands when it comes to safety against earth quake. Primarily, various seismic parameters in BNBC 2010 (draft) have been studied and compared with that of BNBC 1993. Later, both 1993 and 2010 edition of BNBC codes have been compared graphically with building codes of other countries such as National Building Code of India 2005 (NBC-India 2005), American Society of Civil Engineering 7-05 (ASCE 7-05). The base shear/weight ratios have been plotted against the height of the building. The investigation in this paper reveals that BNBC 1993 has the least base shear among all the codes. Factored Base shear values of BNBC 2010 are found to have increased significantly than that of BNBC 1993 for low rise buildings (≤20 m) around the country than its predecessor. Despite revision of the code, BNBC 2010 (draft) still suggests less base shear values when compared to the Indian and American code. Therefore, this increase in factor of safety against the earthquake imposed by the proposed BNBC 2010 code by suggesting higher values of base shear is appreciable.

  3. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  4. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  5. TRAC-PF1/MOD3 calculations of Savannah River Laboratory Rig FA single-annulus heated experiments

    International Nuclear Information System (INIS)

    Fischer, S.R.; McDaniel, C.K.

    1992-01-01

    This paper presents the results of TRAC-PF1/MOD3 benchmarks of the Rig FA experiments performed at the Savannah River Laboratory to simulate prototypic reactor fuel assembly behavior over a range of fluid conditions typical of the emergency cooling system (ECS) phase of a loss-of-coolant accident (LOCA). The primary purpose of this work was to use the SRL Rig FA tests to qualify the TRAC-PF1/MOD3 computer code and models for computing Mark-22 fuel assembly LOCA/ECS power limits. This qualification effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to independently confirm power limits for the Savannah River Site K Reactor. The results of this benchmark effort as discussed in this paper demonstrate that TRAC/PF1/MOD3 coupled with proper modeling is capable of simulating thermal-hydraulic phenomena typical of that encountered in Mark-22 fuel assembly during LOCA/ECS conditions

  6. A code for structural analysis of fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-08-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each tubular structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equilavent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  8. 76 FR 57982 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-09-19

    ... DEPARTMENT OF ENERGY Office of Energy Efficiency and Renewable Energy [Docket No. EERE-2011-BT-BC-0046] Building Energy Codes Cost Analysis Correction In notice document 2011-23236 beginning on page...-23236 Filed 9-16-11; 8:45 am] BILLING CODE 1505-01-P ...

  9. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  10. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)

  11. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).

  12. User's manual for seismic analysis code 'SONATINA-2V'

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Iyoku, Tatsuo

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  13. Implementation and Performance of FPGA based track fitting for the Atlas Fast TracKer

    CERN Document Server

    Zou, Rui; The ATLAS collaboration

    2018-01-01

    The Fast TracKer (FTK) within the ATLAS trigger system provides global track reconstruction for all events passing the ATLAS Level 1 trigger by dividing the detector into parallel processing pipelines that implement pattern matching in custom integrated circuits and data routing, reduction, and parameter extraction in FPGAs. In this presentation we will describe the implementation of a critical component of the system which does partial track fitting using a method based on a principal component analysis at a rate of greater than 1 fit per 10 ps, system-wide, to reduce the output of the pattern matching. Firmware design, timing performance and preliminary results will be discussed.

  14. Production of analysis code for 'JOYO' dosimetry experiment

    International Nuclear Information System (INIS)

    Sasaki, Makoto; Nakazawa, Masaharu.

    1981-01-01

    As part of the measurement and analysis plan for the Dosimetry Experiment at the ''JOYO'' experimental fast reactor, neutron flux spectra analysis is performed using the NEUPAC (Neutron Unfolding Code Package) computer program. The code calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils. The NEUPAC code is based on the J1-type unfolding method, and the estimated neutron flux spectra is obtained as its solution. The program is able to determine the integral quantities and their sensitivities, together with an error estimate of the unfolded spectra and integral quantities. The code also performs a chi-square test of the input/output data, and contains many options for the calculational routines. This report presents the analytic theory, the program algorithms, and a description of the functions and use of the NEUPAC code. (author)

  15. Thermohydraulic analysis of nuclear power plant accidents by computer codes

    International Nuclear Information System (INIS)

    Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.

    1982-01-01

    RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)

  16. GPS-based Microenvironment Tracker (MicroTrac) Model to ...

    Science.gov (United States)

    A critical aspect of air pollution exposure assessment is the estimation of the time spent by individuals in various microenvironments (ME). Accounting for the time spent in different ME with different pollutant concentrations can reduce exposure misclassifications, while failure to do so can add uncertainty and bias to risk estimates. In this study, a classification model, called MicroTrac, was developed to estimate time of day and duration spent in eight ME (indoors and outdoors at home, work, school; inside vehicles; other locations) from global positioning system (GPS) data and geocoded building boundaries. Based on a panel study, MicroTrac estimates were compared to 24 h diary data from 7 participants on workdays and 2 participants on nonworkdays, with corresponding GPS data and building boundaries of home, school, and work. MicroTrac correctly classified the ME for 99.5% of the daily time spent by the participants. The capability of MicroTrac could help to reduce the time-location uncertainty in air pollution exposure models and exposure metrics for individuals in health studies. The National Exposure Research Laboratory’s (NERL’s) Human Exposure and Atmospheric Sciences Division (HEASD) conducts research in support of EPA’s mission to protect human health and the environment. HEASD’s research program supports Goal 1 (Clean Air) and Goal 4 (Healthy People) of EPA’s strategic plan. More specifically, our division conducts research to characterize

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  18. Sensitivity analysis of the RESRAD, a dose assessment code

    International Nuclear Information System (INIS)

    Yu, C.; Cheng, J.J.; Zielen, A.J.

    1991-01-01

    The RESRAD code is a pathway analysis code that is designed to calculate radiation doses and derive soil cleanup criteria for the US Department of Energy's environmental restoration and waste management program. the RESRAD code uses various pathway and consumption-rate parameters such as soil properties and food ingestion rates in performing such calculations and derivations. As with any predictive model, the accuracy of the predictions depends on the accuracy of the input parameters. This paper summarizes the results of a sensitivity analysis of RESRAD input parameters. Three methods were used to perform the sensitivity analysis: (1) Gradient Enhanced Software System (GRESS) sensitivity analysis software package developed at oak Ridge National Laboratory; (2) direct perturbation of input parameters; and (3) built-in graphic package that shows parameter sensitivities while the RESRAD code is operational

  19. PACC information management code for common cause failures analysis

    International Nuclear Information System (INIS)

    Ortega Prieto, P.; Garcia Gay, J.; Mira McWilliams, J.

    1987-01-01

    The purpose of this paper is to present the PACC code, which, through an adequate data management, makes the task of computerized common-mode failure analysis easier. PACC processes and generates information in order to carry out the corresponding qualitative analysis, by means of the boolean technique of transformation of variables, and the quantitative analysis either using one of several parametric methods or a direct data-base. As far as the qualitative analysis is concerned, the code creates several functional forms for the transformation equations according to the user's choice. These equations are subsequently processed by boolean manipulation codes, such as SETS. The quantitative calculations of the code can be carried out in two different ways: either starting from a common cause data-base, or through parametric methods, such as the Binomial Failure Rate Method, the Basic Parameters Method or the Multiple Greek Letter Method, among others. (orig.)

  20. Adaptable Value-Set Analysis for Low-Level Code

    OpenAIRE

    Brauer, Jörg; Hansen, René Rydhof; Kowalewski, Stefan; Larsen, Kim G.; Olesen, Mads Chr.

    2012-01-01

    This paper presents a framework for binary code analysis that uses only SAT-based algorithms. Within the framework, incremental SAT solving is used to perform a form of weakly relational value-set analysis in a novel way, connecting the expressiveness of the value sets to computational complexity. Another key feature of our framework is that it translates the semantics of binary code into an intermediate representation. This allows for a straightforward translation of the program semantics in...

  1. Improvement of QR Code Recognition Based on Pillbox Filter Analysis

    Directory of Open Access Journals (Sweden)

    Jia-Shing Sheu

    2013-04-01

    Full Text Available The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the contents in the QR code image. Many studies have used the pillbox filter (circular averaging filter method to simulate an out-of-focus image. This method is also used in this investigation to improve the recognition of a captured QR code image. A blurred QR code image is separated into nine levels. In the experiment, four different quantitative approaches are used to reconstruct and decode an out-of-focus QR code image. These nine reconstructed QR code images using methods are then compared. The final experimental results indicate improvements in identification.

  2. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  3. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  4. A compendium of computer codes in fault tree analysis

    International Nuclear Information System (INIS)

    Lydell, B.

    1981-03-01

    In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)

  5. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

  6. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

  7. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  8. Automated uncertainty analysis methods in the FRAP computer codes

    International Nuclear Information System (INIS)

    Peck, S.O.

    1980-01-01

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

  9. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  10. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  11. A plan for the modification and assessment of TRAC-PF1/MOD2 for use in analyzing CANDU 3 transient thermal-hydraulic phenomena

    International Nuclear Information System (INIS)

    Siebe, D.A.; Boyack, B.E.; Giguere, P.T.

    1994-11-01

    This report presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-hydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PF1/MOD2 (TRAC) computer code needed to adequately predict CANDU 3 transient thermal-hydraulic phenomena, and identify an assessment program to verify the ability of TRAC, when modified, to predict these phenomena. This work builds on analyses and recommendations produced by the Idaho National Engineering Laboratory (INEL). To identify the thermal-hydraulic phenomena, a large-break loss-of-coolant accident simulation, performed as part of earlier work by INEL with an Atomic Energy of Canada, Limited (AECL) thermal-hydraulic computer code (CATHENA), was analyzed in detail. Other accident scenarios were examined for additional phenomena. A group of Los Alamos National Laboratory reactor thermal-hydraulics experts ranked the phenomena to produce a preliminary phenomena identification and ranking table (PIRT). The preliminary nature of the PIRT was a result of a lack of direct expertise with the unique processes and phenomena of the CANDU 3. Nonetheless, this PIRT provided an adequate foundation for planning a program of code modifications. We believe that this PIRT captured the most important phenomena and that refinements to the PIRT will mainly produce clarification of the relative importance (ranking) of phenomena. A plan for code modifications was developed based on this PIRT and on information about the modeling methodologies for CANDU-specific phenomena used in AECL codes. AECL thermal-hydraulic test facilities and programs were reviewed and the information used in developing an assessment plan to ensure that TRAC-PF1/MOD2, when modified, will adequately predict CANDU 3 phenomena

  12. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

    2006-08-01

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

  13. Guide to Using Onionskin Analysis Code (U)

    Energy Technology Data Exchange (ETDEWEB)

    Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group; Morzinski, Jerome Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group

    2016-09-15

    This document is a guide to using R-code written for the purpose of analyzing onionskin experiments. We expect the user to be very familiar with statistical methods and the R programming language. For more details about onionskin experiments and the statistical methods mentioned in this document see Storlie, Fugate, et al. (2013). Engineers at LANL experiment with detonators and high explosives to assess performance. The experimental unit, called an onionskin, is a hemisphere consisting of a detonator and a booster pellet surrounded by explosive material. When the detonator explodes, a streak camera mounted above the pole of the hemisphere records when the shock wave arrives at the surface. The output from the camera is a two-dimensional image that is transformed into a curve that shows the arrival time as a function of polar angle. The statistical challenge is to characterize a baseline population of arrival time curves and to compare the baseline curves to curves from a new, so-called, test series. The hope is that the new test series of curves is statistically similar to the baseline population.

  14. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Glass, R.E.

    1989-01-01

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

  15. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

    Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

    1988-03-01

    Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

  16. Color Coding of Circuit Quantities in Introductory Circuit Analysis Instruction

    Science.gov (United States)

    Reisslein, Jana; Johnson, Amy M.; Reisslein, Martin

    2015-01-01

    Learning the analysis of electrical circuits represented by circuit diagrams is often challenging for novice students. An open research question in electrical circuit analysis instruction is whether color coding of the mathematical symbols (variables) that denote electrical quantities can improve circuit analysis learning. The present study…

  17. Multi-pressure boiler thermodynamics analysis code

    International Nuclear Information System (INIS)

    Lorenzoni, G.

    1992-01-01

    A new method and the relative FORTRAN program for the thermodynamics design analysis of a multipressure boiler are reported. This method permits the thermodynamics design optimization with regard to total exergy production and a preliminary costs

  18. Analysis of Iterated Hard Decision Decoding of Product Codes with Reed-Solomon Component Codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2007-01-01

    Products of Reed-Solomon codes are important in applications because they offer a combination of large blocks, low decoding complexity, and good performance. A recent result on random graphs can be used to show that with high probability a large number of errors can be corrected by iterating...... minimum distance decoding. We present an analysis related to density evolution which gives the exact asymptotic value of the decoding threshold and also provides a closed form approximation to the distribution of errors in each step of the decoding of finite length codes....

  19. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  20. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  1. Development and validation of sodium fire analysis code ASSCOPS

    International Nuclear Information System (INIS)

    Ohno, Shuji

    2001-01-01

    A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)

  2. A static analysis tool set for assembler code verification

    International Nuclear Information System (INIS)

    Dhodapkar, S.D.; Bhattacharjee, A.K.; Sen, Gopa

    1991-01-01

    Software Verification and Validation (V and V) is an important step in assuring reliability and quality of the software. The verification of program source code forms an important part of the overall V and V activity. The static analysis tools described here are useful in verification of assembler code. The tool set consists of static analysers for Intel 8086 and Motorola 68000 assembly language programs. The analysers examine the program source code and generate information about control flow within the program modules, unreachable code, well-formation of modules, call dependency between modules etc. The analysis of loops detects unstructured loops and syntactically infinite loops. Software metrics relating to size and structural complexity are also computed. This report describes the salient features of the design, implementation and the user interface of the tool set. The outputs generated by the analyser are explained using examples taken from some projects analysed by this tool set. (author). 7 refs., 17 figs

  3. Fault tree analysis. Implementation of the WAM-codes

    International Nuclear Information System (INIS)

    Bento, J.P.; Poern, K.

    1979-07-01

    The report describes work going on at Studsvik at the implementation of the WAM code package for fault tree analysis. These codes originally developed under EPRI contract by Sciences Applications Inc, allow, in contrast with other fault tree codes, all Boolean operations, thus allowing modeling of ''NOT'' conditions and dependent components. To concretize the implementation of these codes, the auxiliary feed-water system of the Swedish BWR Oskarshamn 2 was chosen for the reliability analysis. For this system, both the mean unavailability and the probability density function of the top event - undesired event - of the system fault tree were calculated, the latter using a Monte-Carlo simulation technique. The present study is the first part of a work performed under contract with the Swedish Nuclear Power Inspectorate. (author)

  4. Code conversion for system design and safety analysis of NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)

  5. TRAC-PF1 analyses of potential pressurized-thermal-shock transients at a Combustion-Engineering PWR

    International Nuclear Information System (INIS)

    Koenig, J.E.; Spriggs, G.D.; Smith, R.C.

    1984-01-01

    Los Alamos is participating in a program to assess the risk of pressurized thermal shock (PTS) to a reactor vessel. Our role is to provide best-estimate thermal-hydraulic analyses of 12 postulated overcooling transients using TRAC-PF1. These transients are hypothetical and include multiple operator/equipment failures. Calvert Cliffs/Unit-1, a Combustion-Engineering plant, is the pressurized water reactor modeled for this study. The utility and the vendor supplied information for the comprehensive TRAC-PF1 model. Secondary and primary breaks from both hot-zero-power and full-power conditions were simulated for 7200 s (2 h). Low bulk temperatures and loop-flow stagnation while the system was at a high pressure were of particular interest for PTS analysis. Three transients produced primary temperatures below 405 K (270 0 F - the NRC screening criterion) with system repressurization. Six transients indicated flow stagnation would occur in one loop but not both. One transient showed flow stagnation might occur in both loops. Oak Ridge National Laboratory will do fracture-mechanics analysis using these TRAC-PF1 results and make the final determination of the risk of PTS

  6. Development and improvement of safety analysis code for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  7. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    Millard, D.; Hoffmann, A.

    1985-02-01

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented

  8. Implementation and assessment of improved models and options in TRAC-BF1

    International Nuclear Information System (INIS)

    Analytis, G.Th.

    1996-07-01

    A summary of modifications and options introduced in TRAC-BF1 is presented and it is shown that the predicting capabilities of the modified version of the core are greatly improved. These changes include the introduction of a different heat transfer package during reflooding, the implementation of a simple single-phase limit procedure for forcing the two phases to acquire the same velocity if one phase disappears, a close assessment of the annular flow interfacial shear correlation, implementation of a simple radiation model which seems to alleviate some numerical-oscillations problems induced by the existing highly complex model. Furthermore, different options were introduced and tested like upwinding some terms of the momentum equations (which seems to solve a number of problems reported in the past), the second upwind scheme for the convective terms of the phasic momentum equations and the implementation and assessment of a completely different annular flow interfacial shear correlation. The modified TRAC-BF1 is assessed against some bottom-flooding separate-effect experiments, a 'benchmark' top flooding simulation as well as against the TLTA test Nr. 6423. In the process of this task, the different options are assessed and discussed and is shown that the predictions of the modified code are physically sound and close 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 for LB-LOCA analyses. (author) 19 figs., 16 refs

  9. Two-dimensional disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.

    1988-08-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

  10. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

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

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1985-01-01

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

  12. PCLOOK: an interactive code for spectral analysis

    International Nuclear Information System (INIS)

    Macchiavelli, A.O.; Tomasi, D.

    1993-01-01

    The present work describes an interactive programme for the analysis of spectra developed to run in a PC platform. PCLOOK has a graphic interface that allows the user to get access to different functions using the mouse or directly typing commands. In this way one can switch to a suitable required environment to manage the histograms reassembling in this way a spectrum calculator.The PCLOOK programme was mainly developed to use in nuclear physics applications, but it is also possible to modify it with relative little effort to adapt it to other applications. It was written in Microsoft's BASIC 7.1 installed in a 33MHz 486 Everex PC. For a proper operation an ordinary VGA display and mouse are needed. The memory requirements depend on the size and number of the user defined spectra; for instance, for twenty 2048 channels spectra the available memory space must be 320 KBytes. (author). 5 figs

  13. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Glass, R.E.

    1989-01-01

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

  14. Integrated severe accident containment analysis with the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

    1985-12-01

    Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

  15. NACOM - a code for sodium spray fire analysis

    International Nuclear Information System (INIS)

    Rao, P.M.; Kannan, S.E.

    2002-01-01

    Full text: In liquid metal fast breeder reactors (LMFBR), leakage of sodium can result in a spray fire. Because of higher burning rates in droplet form combustion of sodium in spray fire, thermal consequences are more severe than that in a sodium pool fire. The code NACOM was developed for the analysis of sodium spray fires in LMFBRs facilities. The code uses the validated model for estimating the falling droplet burning rates in pre-ignition and vapour phase combustion stages. It uses a distribution system to generate the droplet groups of different diameters that represent the spray. The code requires about 20 input parameters like sodium leak rates, sodium temperature, initial cell conditions like oxygen concentration, temperature and dimensions. NACOM is a validated code based on experiments with sodium inventory up to 650 kg in 0 to 21 % O 2 atmospheres. The paper brings out the salient features of the code along with the sensitivity analysis of the main input parameters like spray volume mean diameter, oxygen concentration etc. based on the results obtained. The limitations of the code and the confidence margins applicable to results obtained are also brought out

  16. Analysis of pipe stress using CAESAR II code

    International Nuclear Information System (INIS)

    Sitandung, Y.B.; Bandriyana, B.

    2002-01-01

    Analysis of this piping stress with the purpose of knowing stress distribution piping system in order to determine pipe supports configuration. As an example of analysis, Gas Exchanger to Warm Separator Line was chosen with, input data was firstly prepared in a document, i.e. piping analysis specification that its content named as pipe characteristics, material properties, operation conditions, guide equipment's and so on. Analysis result such as stress, load, displacement and the use support type were verified based on requirements in the code, standard, and regularities were suitable with piping system condition analyzed. As the proof that piping system is in safety condition, it can be indicated from analysis results (actual loads) which still under allowable load. From the analysis steps that have been done CAESAR II code fulfill requirements to be used as a tool of piping stress analysis as well as nuclear and non nuclear installation piping system

  17. NORTICA—a new code for cyclotron analysis

    Science.gov (United States)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-12-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER [1] developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state.

  18. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  19. Two-phase flow characteristics analysis code: MINCS

    International Nuclear Information System (INIS)

    Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

    1992-03-01

    Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. TRAC-BDl/MOD1 post-dryout wall heat transfer

    International Nuclear Information System (INIS)

    Shumway, R.W.

    1984-01-01

    A comparison of TRAC-BWR heat transfer package with 766 data points is presented. On the average, TRAC-BWR provides a better prediction of the data compared to any single correlation although there is still a large scatter in TRAC-BWR prediction. Regarding any potential changes in the TRAC-BD1/MOD1 wall heat transfer package, it is concluded that no significant improvement in the film boiling area can be made until data with better measurements are obtained and analyzed. Specifically, data is needed which has a wide range of accurately measured void fractions. Heated tube data is also needed which addresses the countercurrent flow transition conditions

  3. SAFARI 2000 FPAR TRAC Data for Mongu, Zambia, 1999-2002

    Data.gov (United States)

    National Aeronautics and Space Administration — Data from the Tracing Architecture and Radiation of Canopies (TRAC) instrument were processed to determine the fraction of intercepted photosynthetically active...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  5. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

    2001-01-01

    Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)

  6. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

  7. Development of a code for the isotopic analysis of Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    To strengthen the national nuclear nonproliferation regime by an establishment of nuclear forensic system, the techniques for nuclear material analysis and the categorization of important domestic nuclear materials are being developed. MGAU and FRAM are commercial software for the isotopic analysis of Uranium by using γ-spectroscopy, but the diversity of detection geometry and some effects - self attenuation, coincidence summing, etc. - suggest an analysis tool under continual improvement and modification. Hence, developing another code for HPGe γ- and x-ray spectrum analysis is started in this study. The analysis of the 87-101 keV region of Uranium spectrum is attempted based on the isotopic responses similar to those developed in MGAU. The code for isotopic analysis of Uranium is started from a fitting.

  8. QR Codes in the Library: Are They Worth the Effort? Analysis of a QR Code Pilot Project

    OpenAIRE

    Wilson, Andrew M.

    2012-01-01

    The literature is filled with potential uses for Quick Response (QR) codes in the library. Setting, but few library QR code projects have publicized usage statistics. A pilot project carried out in the Eda Kuhn Loeb Music Library of the Harvard College Library sought to determine whether library patrons actually understand and use QR codes. Results and analysis of the pilot project are provided, attempting to answer the question as to whether QR codes are worth the effort for libraries.

  9. TRAC, a collaborative computer tool for tracer-test interpretation

    Directory of Open Access Journals (Sweden)

    Fécamp C.

    2013-05-01

    Full Text Available Artificial tracer tests are widely used by consulting engineers for demonstrating water circulation, proving the existence of leakage, or estimating groundwater velocity. However, the interpretation of such tests is often very basic, with the result that decision makers and professionals commonly face unreliable results through hasty and empirical interpretation. There is thus an increasing need for a reliable interpretation tool, compatible with the latest operating systems and available in several languages. BRGM, the French Geological Survey, has developed a project together with hydrogeologists from various other organizations to build software assembling several analytical solutions in order to comply with various field contexts. This computer program, called TRAC, is very light and simple, allowing the user to add his own analytical solution if the formula is not yet included. It aims at collaborative improvement by sharing the tool and the solutions. TRAC can be used for interpreting data recovered from a tracer test as well as for simulating the transport of a tracer in the saturated zone (for the time being. Calibration of a site operation is based on considering the hydrodynamic and hydrodispersive features of groundwater flow as well as the amount, nature and injection mode of the artificial tracer. The software is available in French, English and Spanish, and the latest version can be downloaded from the web site http://trac.brgm.fr.

  10. Computer codes for the analysis of flask impact problems

    International Nuclear Information System (INIS)

    Neilson, A.J.

    1984-09-01

    This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)

  11. Java Source Code Analysis for API Migration to Embedded Systems

    Energy Technology Data Exchange (ETDEWEB)

    Winter, Victor [Univ. of Nebraska, Omaha, NE (United States); McCoy, James A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Guerrero, Jonathan [Univ. of Nebraska, Omaha, NE (United States); Reinke, Carl Werner [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perry, James Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    Embedded systems form an integral part of our technological infrastructure and oftentimes play a complex and critical role within larger systems. From the perspective of reliability, security, and safety, strong arguments can be made favoring the use of Java over C in such systems. In part, this argument is based on the assumption that suitable subsets of Java’s APIs and extension libraries are available to embedded software developers. In practice, a number of Java-based embedded processors do not support the full features of the JVM. For such processors, source code migration is a mechanism by which key abstractions offered by APIs and extension libraries can made available to embedded software developers. The analysis required for Java source code-level library migration is based on the ability to correctly resolve element references to their corresponding element declarations. A key challenge in this setting is how to perform analysis for incomplete source-code bases (e.g., subsets of libraries) from which types and packages have been omitted. This article formalizes an approach that can be used to extend code bases targeted for migration in such a manner that the threats associated the analysis of incomplete code bases are eliminated.

  12. Development of the vacuum system pressure responce analysis code PRAC

    International Nuclear Information System (INIS)

    Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

    1985-03-01

    In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

  13. Comparison study of inelastic analysis codes for high temperature structure

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Lee, H. Y.; Park, C. K.; Geon, G. P.; Lee, J. H

    2004-02-01

    LMR high temperature structures subjected to operating and transient loadings may exhibit very complex deformation behaviors due to the use of ductile material such as 316SS and the systematic analysis technology of high temperature structure for reliable safety assessment is essential. In this project, comparative study with developed inelastic analysis program NONSTA and the existing analysis codes was performed applying various types of loading including non-proportional loading. The performance of NONSTA was confirmed and the effect of inelastic constants on the analysis result was analyzed. Also, the applicability of the inelastic analysis was enlarged as a result of applying both the developed program and the existing codes to the analyses of the enhanced creep behavior and the elastic follow-up behavior of high temperature structures and the necessary items for improvements were deduced. Further studies on the improvement of NONSTA program and the decision of the proper values of inelastic constants are necessary.

  14. Easy web interfaces to IDL code for NSTX Data Analysis

    International Nuclear Information System (INIS)

    Davis, W.M.

    2012-01-01

    Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

  15. [Bioethical analysis of the Brazilian Dentistry Code of Ethics].

    Science.gov (United States)

    Pyrrho, Monique; do Prado, Mauro Machado; Cordón, Jorge; Garrafa, Volnei

    2009-01-01

    The Brazilian Dentistry Code of Ethics (DCE), Resolution CFO-71 from May 2006, is an instrument created to guide dentists' behavior in relation to the ethical aspects of professional practice. The purpose of the study is to analyze the above mentioned code comparing the deontological and bioethical focuses. In order to do so, an interpretative analysis of the code and of twelve selected texts was made. Six of the texts were about bioethics and six on deontology, and the analysis was made through the methodological classification of the context units, textual paragraphs and items from the code in the following categories: the referentials of bioethical principlism--autonomy, beneficence, nonmaleficence and justice -, technical aspects and moral virtues related to the profession. Together the four principles represented 22.9%, 39.8% and 54.2% of the content of the DCE, of the deontological texts and of the bioethical texts respectively. In the DCE, 42% of the items referred to virtues, 40.2% were associated to technical aspects and just 22.9% referred to principles. The virtues related to the professionals and the technical aspects together amounted to 70.1% of the code. Instead of focusing on the patient as the subject of the process of oral health care, the DCE focuses on the professional, and it is predominantly turned to legalistic and corporate aspects.

  16. Development of a subchannel analysis code MATRA (Ver. α)

    International Nuclear Information System (INIS)

    Yoo, Y. J.; Hwang, D. H.

    1998-04-01

    A subchannel analysis code MATRA-α, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-α has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-α. In addition, we compared the predictions of MATRA-α with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-α. All the results revealed that the prediction of MATRA-α were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs

  17. ISODEP, A Fuel Depletion Analysis Code for Predicting Isotopic ...

    African Journals Online (AJOL)

    The trend of results was found to be consistent with those obtained by analytical and other numerical methods. Discovery and Innovation Vol. 13 no. 3/4 December (2001) pp. 184-195. KEY WORDS: depletion analysis, code, research reactor, simultaneous equations, decay of nuclides, radionuclitides, isotope. Résumé

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

    International Nuclear Information System (INIS)

    Mulas, J.; Queral, C.; Collazo, I.; Concejal, A.; Burbano, N.; Lopez Lechas, A.; Tarrega, I.

    2002-01-01

    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)

  19. PBF severe fuel damage program: results and comparison to analysis

    International Nuclear Information System (INIS)

    McDonald, P.E.; Buescher, B.J.; Gruen, G.E.; Hobbins, R.R.; McCardell, R.K.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  20. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  1. FARO and KROTOS code simulation and analysis at JRC Ispra

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)

    1998-01-01

    The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)

  2. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

  3. Integrated computer codes for nuclear power plant severe accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

  4. Room Heat-Up Analysis with GOTHIC code

    International Nuclear Information System (INIS)

    Jimenez, G.; Olza, J. M.

    2010-01-01

    The GOTHIC T M computer code is a state-of-the art program for modeling multiphase, multicomponent fluid flow. GOTHIC is rapidly becoming the industry-standard code for performing both containment design basis accident (DBA) analyses and analyses to support equipment qualification. GOTHIC has a flexible nodding structure that allows both lumped parameter and 3-D modeling capabilities. Multidimensional analysis capabilities greatly enhance the study of noncondensable gases and stratification and permit the calculation of flow field details within any given volume.

  5. RADTRAN 5: A computer code for transportation risk analysis

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Kanipe, F.L.

    1991-01-01

    RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air

  6. Organization of Risk Analysis Codes for Living Evaluations (ORACLE)

    International Nuclear Information System (INIS)

    Batt, D.L.; MacDonald, P.E.; Sattison, M.B.; Vesely, E.

    1987-01-01

    ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)

  7. Probabilistic analysis of crack containing structures with the PARIS code

    International Nuclear Information System (INIS)

    Brueckner-Foit, A.

    1987-10-01

    The basic features of the PARIS code which has been developed for the calculation of failure probabilities of crack containing structures are explained. An important issue in the reliability analysis of cracked components is the probabilistic leak-before-break behaviour. Formulae for the leak and break probabilities are derived and it is shown how a leak detection system influences the results. An example taken from nuclear applications illustrates the details of the probabilistic leak-before-break analysis. (orig.) [de

  8. Development and assessment of best estimate integrated safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

    2007-03-15

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

  9. Development and assessment of best estimate integrated safety analysis code

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu

    2007-03-01

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published

  10. Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc. Sungnam, (Korea, Republic of); Song, Dong Soo; Byun, Choong Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences.

  11. Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

    International Nuclear Information System (INIS)

    Seo, Kwi Hyun; Song, Wan Jung; Song, Dong Soo; Byun, Choong Sup

    2006-01-01

    The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences

  12. Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code

    International Nuclear Information System (INIS)

    Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun

    2012-01-01

    SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment

  13. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  14. Analysis of the anisotropy effects with the AZTRAN code

    International Nuclear Information System (INIS)

    Xolocostli, V.; Vargas, S.; Gomez, A.; Del Valle, E.

    2017-09-01

    Among the improvements that are made for the deterministic codes with which nuclear reactors are analyzed, is the implementation of the dispersion anisotropic dispersion section, which can obtain better results. With the current computing technology is possible to carry out these implementations, since the computation time is no longer a considerable problem as in the past. In this paper we analyze some effects of anisotropy in the AZTRAN code, a code that solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multigroup technique, the nodal method RTN-0 and ordered discrete, which is part of the AZTLAN platform for analysis of nuclear reactors, which is currently under development. The implementation of the anisotropy in the AZTRAN code is one of the latest improvements that have been made to the code, leading to different tests and analyzes regarding the anisotropic dispersion, some as a test with homogeneous fuel assemblies. In the case presented here, the benchmark problem of a fuel assembly type BWR is analyzed, which is part of the Benchmark problem suite for reactor physics study of LWR next generation fuels, proposed by the Committee on Reactor Physics organized by the Japan Atomic Energy Research Institute (JAERI). In this problem the behavior of the infinite multiplication factor (k inf ) is analyzed, as well as the behavior of using odd and even anisotropy approximation with respect to the symmetry in the radial power of the assembly. (Author)

  15. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

    As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

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

    Science.gov (United States)

    Savage, B.; Snoke, J. A.

    2017-12-01

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

  17. 78 FR 24313 - Proposed Collection; Comment Request for Tip Reporting Alternative Commitment Agreement (TRAC...

    Science.gov (United States)

    2013-04-24

    ... Reporting Alternative Commitment Agreement (TRAC) For Use in Industries Other Than the Food and Beverage... Use in Industries Other than the Food and Beverage Industry and The Cosmetology and Barber Industry... Commitment Agreement (TRAC) For Use in Industries Other than the Food and Beverage Industry and The...

  18. 75 FR 11226 - Proposed Collection; Comment Request for Tip Reporting Alternative Commitment Agreement (TRAC...

    Science.gov (United States)

    2010-03-10

    ... Reporting Alternative Commitment Agreement (TRAC) for Use in the Food and Beverage Industry AGENCY: Internal... Reporting Alternative Commitment Agreement (TRAC) for Use in the Food and Beverage Industry. DATES: Written... in the Food and Beverage Industry. OMB Number: 1545-1549. Abstract: Announcement 2000-22, 2000-19 I.R...

  19. 75 FR 11225 - Proposed Collection; Comment Request for Tip Reporting Alternative Commitment Agreement (TRAC...

    Science.gov (United States)

    2010-03-10

    ... Reporting Alternative Commitment Agreement (TRAC) for Use in Industries Other Than the Food and Beverage... Use in Industries Other Than the Food and Beverage Industry and the Cosmetology and Barber Industry... Alternative Commitment Agreement (TRAC) for Use in Industries Other Than the Food and Beverage Industry and...

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

    Directory of Open Access Journals (Sweden)

    Takumi Murata

    2018-01-01

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

  1. Qualification of ARROTTA code for LWR accident analysis

    International Nuclear Information System (INIS)

    Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

    2004-01-01

    This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

  2. Code portability and data management considerations in the SAS3D LMFBR accident-analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1981-01-01

    The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available

  3. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

  4. THYDE-P2 code: RCS (reactor-coolant system) analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

    1986-12-01

    THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

    2013-10-01

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

  7. Computer code for qualitative analysis of gamma-ray spectra

    International Nuclear Information System (INIS)

    Yule, H.P.

    1979-01-01

    Computer code QLN1 provides complete analysis of gamma-ray spectra observed with Ge(Li) detectors and is used at both the National Bureau of Standards and the Environmental Protection Agency. It locates peaks, resolves multiplets, identifies component radioisotopes, and computes quantitative results. The qualitative-analysis (or component identification) algorithms feature thorough, self-correcting steps which provide accurate isotope identification in spite of errors in peak centroids, energy calibration, and other typical problems. The qualitative-analysis algorithm is described in this paper

  8. A GFR benchmark comparison of transient analysis codes based on the ETDR concept

    International Nuclear Information System (INIS)

    Bubelis, E.; Coddington, P.; Castelliti, D.; Dor, I.; Fouillet, C.; Geus, E. de; Marshall, T.D.; Van Rooijen, W.; Schikorr, M.; Stainsby, R.

    2007-01-01

    A GFR (Gas-cooled Fast Reactor) transient benchmark study was performed to investigate the ability of different code systems to calculate the transition in the core heat removal from the main circuit forced flow to natural circulation cooling using the Decay Heat Removal (DHR) system. This benchmark is based on a main blower failure in the Experimental Technology Demonstration Reactor (ETDR) with reactor scram. The codes taking part into the benchmark are: RELAP5, TRAC/AAA, CATHARE, SIM-ADS, MANTA and SPECTRA. For comparison purposes the benchmark was divided into several stages: the initial steady-state solution, the main blower flow run-down, the opening of the DHR loop and the transition to natural circulation and finally the 'quasi' steady heat removal from the core by the DHR system. The results submitted by the participants showed that all the codes gave consistent results for all four stages of the benchmark. In the steady-state the calculations revealed some differences in the clad and fuel temperatures, the core and main loop pressure drops and in the total Helium mass inventory. Also some disagreements were observed in the Helium and water flow rates in the DHR loop during the final natural circulation stage. Good agreement was observed for the total main blower flow rate and Helium temperature rise in the core, as well as for the Helium inlet temperature into the core. In order to understand the reason for the differences in the initial 'blind' calculations a second round of calculations was performed using a more precise set of boundary conditions

  9. Clinical accuracy of ExacTrac intracranial frameless stereotactic system

    International Nuclear Information System (INIS)

    Ackerly, T.; Lancaster, C. M.; Geso, M.; Roxby, K. J.

    2011-01-01

    Purpose: In this paper, the authors assess the accuracy of the Brainlab ExacTrac system for frameless intracranial stereotactic treatments in clinical practice. Methods: They recorded couch angle and image fusion results (comprising lateral, longitudinal, and vertical shifts, and rotation corrections about these axes) for 109 stereotactic radiosurgery and 166 stereotactic radiotherapy patient treatments. Frameless stereotactic treatments involve iterative 6D image fusion corrections applied until the results conform to customizable pass criteria, theirs being 0.7 mm and 0.5 deg. for each axis. The planning CT slice thickness was 1.25 mm. It has been reported in the literature that the CT slices' thickness impacts the accuracy of localization to bony anatomy. The principle of invariance with respect to patient orientation was used to determine spatial accuracy. Results: The data for radiosurgery comprised 927 image pairs, of which 532 passed (pass ratio of 57.4%). The data for radiotherapy comprised 15983 image pairs, of which 10 050 passed (pass ratio of 62.9%). For stereotactic radiotherapy, the combined uncertainty of ExacTrac calibration, image fusion, and intrafraction motion was (95% confidence interval) 0.290-0.302 and 0.306-0.319 mm in the longitudinal and lateral axes, respectively. The combined uncertainty of image fusion and intrafraction motion in the anterior-posterior coordinates was 0.174-0.182 mm. For stereotactic radiosurgery, the equivalent ranges are 0.323-0.393, 0.337-0.409, and 0.231-0.281 mm. The overall spatial accuracy was 1.24 mm for stereotactic radiotherapy (SRT) and 1.35 mm for stereotactic radiosurgery (SRS). Conclusions: The ExacTrac intracranial frameless stereotactic system spatial accuracy is adequate for clinical practice, and with the same pass criteria, SRT is more accurate than SRS. They now use frameless stereotaxy exclusively at their center.

  10. Moderator 3-D Thermalhydraulic Analysis Using MODTURCCLAS Code

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Yoon, Churl; Park, Joo Hwan

    2008-12-01

    For the moderator subcooling analysis of the refurbished Wolsong NPP (Nuclear Power Plant) Unit 1, the 3-D moderator thermalhydraulic calculations were preliminarily conducted from September to October in 2008, using the preliminary thermalhydraulic analysis results and the MODTURC C LAS (MODerator TURbulent Circulation Co-Located Advanced Solutions) Ver.2.9-IST, which was developed and validated by OPG (Ontario Power Generation) in Canada. The present report consists of the steady-state calculation and transient calculation. First the grid structure, main input parameters, and boundary conditions needed for the steady-state calculation are produced and the steady-state results are obtained. These steady-state results are used for the initial conditions for the transient analysis during a LOCA. The moderator analysis results during the steady-state calculation show the quasy-steady state behavior, because the thermalhydraulic behavior are fluctuating although all boundary conditions are constant. In the transient calculations, based on the present thermalhydraulic analysis results, 3-D thermalhydraulic behavior and moderator subcooling are predicted for the accident scenarios of reactor inlet header 35% and 40% breaks, outlet header 100% break, and pump suction 80% break, subsequent with loss of Class IV power. In the previous moderator analysis for Wolsong NPP Unit 2,3,4 a PHOENICS code was used, which is different from the MODTURC C LAS code used for the analysis of Wolsong NPP Unit 1. However, the moderator subcooling analysis results by these two codes are qualitatively similar. The minimum subcooling for RIH 40% break of Wolsong NPP Unit 1 is 17 .deg. C which is larger than 13 .deg. C for RIH 35% break of Wolsong NPP Unit 2,3,4. Therefore, it is concluded that the refurbished Wolsong NPP Unit 1 satisfies the channel integrity criteria based on the higher subcooling margin compared with that of Wolsong NPP Unit 2,3,4

  11. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  12. Source Code Analysis Laboratory (SCALe) for Energy Delivery Systems

    Science.gov (United States)

    2010-12-01

    technical competence for the type of tests and calibrations SCALe undertakes. Testing and calibration laboratories that comply with ISO / IEC 17025 ...and exec t [ ISO / IEC 2005]. f a software system indicates that the SCALe analysis di by a CERT secure coding standard. Successful conforma antees that...to be more secure than non- systems. However, no study has yet been performed to p t ssment in accordance with ISO / IEC 17000: “a demonstr g to a

  13. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel.

  14. LACEwING: A New Moving Group Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Riedel, Adric R. [Department of Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Blunt, Sarah C.; Faherty, Jacqueline K. [Department of Astrophysics, American Museum of Natural History, New York, NY 10024 (United States); Lambrides, Erini L. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Rice, Emily L. [Department of Engineering Science and Physics, The College of Staten Island, Staten Island, NY 10314 (United States); Cruz, Kelle L., E-mail: arr@astro.caltech.edu [Department of Physics and Astronomy, Hunter College, New York, NY 10065 (United States)

    2017-03-01

    We present a new nearby young moving group (NYMG) kinematic membership analysis code, LocAting Constituent mEmbers In Nearby Groups (LACEwING), a new Catalog of Suspected Nearby Young Stars, a new list of bona fide members of moving groups, and a kinematic traceback code. LACEwING is a convergence-style algorithm with carefully vetted membership statistics based on a large numerical simulation of the Solar Neighborhood. Given spatial and kinematic information on stars, LACEwING calculates membership probabilities in 13 NYMGs and three open clusters within 100 pc. In addition to describing the inputs, methods, and products of the code, we provide comparisons of LACEwING to other popular kinematic moving group membership identification codes. As a proof of concept, we use LACEwING to reconsider the membership of 930 stellar systems in the Solar Neighborhood (within 100 pc) that have reported measurable lithium equivalent widths. We quantify the evidence in support of a population of young stars not attached to any NYMGs, which is a possible sign of new as-yet-undiscovered groups or of a field population of young stars.

  15. SHEAN (Simplified Human Error Analysis code) and automated THERP

    International Nuclear Information System (INIS)

    Wilson, J.R.

    1993-01-01

    One of the most widely used human error analysis tools is THERP (Technique for Human Error Rate Prediction). Unfortunately, this tool has disadvantages. The Nuclear Regulatory Commission, realizing these drawbacks, commissioned Dr. Swain, the author of THERP, to create a simpler, more consistent tool for deriving human error rates. That effort produced the Accident Sequence Evaluation Program Human Reliability Analysis Procedure (ASEP), which is more conservative than THERP, but a valuable screening tool. ASEP involves answering simple questions about the scenario in question, and then looking up the appropriate human error rate in the indicated table (THERP also uses look-up tables, but four times as many). The advantages of ASEP are that human factors expertise is not required, and the training to use the method is minimal. Although not originally envisioned by Dr. Swain, the ASEP approach actually begs to be computerized. That WINCO did, calling the code SHEAN, for Simplified Human Error ANalysis. The code was done in TURBO Basic for IBM or IBM-compatible MS-DOS, for fast execution. WINCO is now in the process of comparing this code against THERP for various scenarios. This report provides a discussion of SHEAN

  16. Assessment of TRAC-PF1/MOD1 against a loss-of-grid transient in Ringhals 4 power plant

    International Nuclear Information System (INIS)

    Sjoberg, A.; Almberger, J.; Sandervag, O.

    1989-07-01

    A loss of grid transient in a three loop Westinghouse PWR has been simulated with the frozen version of TRAC-PF1/MOD1 computer code. The results reveal the capability of the code to qualitatively predict the different pertinent phenomena and the data comparison was quite encouraging. Accurate predictions of the system response required careful determination of the boundary conditions simulating the turbine governor valves and steam dump valves behaviour. An explicit modeling of the steam generator internals was also found to be important for the results. It was also revealed that the pressurizer system including spray and heaters and their operation should be modeled in some detail for proper response. 4 refs., 18 figs

  17. Development of the SCHAMBETA code for scoping analysis of HCDA

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Hahn, D. H

    2000-06-01

    A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  19. PBF Severe Fuel-Damage Program: results and comparison to analysis

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Buescher, B.J.; Hobbins, R.R.; McCardell, R.K.; Gruen, G.E.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel-damage research program in the Power Burst Facility (PBF) to investigate fuel-rod and core response, and fission-product and hydrogen release and transport under degraded-core-cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  20. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

    THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)

  1. Development of a neutronic analysis code using data from Monju

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.

    2015-01-01

    In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (author)

  2. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  3. Probabilistic evaluations for CANTUP computer code analysis improvement

    International Nuclear Information System (INIS)

    Florea, S.; Pavelescu, M.

    2004-01-01

    Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

  4. TRAWA, a transient analysis code for water reactions

    International Nuclear Information System (INIS)

    Rajamaeki, M.

    1976-06-01

    TRAWA is a transient analysis code for water reactors. It solves the two-group neutron diffusion equations simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. At most one-dimensional submodels are used. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermohydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channel and risers with two-phase flow and of pump lines with incompressible flow. Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. Various transients can be calculated by applying external disturbances. The code is extensively supplied by input and output capabilities. TRAWA is written in FORTRAN V for UNIVAC 1108 computer. (author)

  5. Beam Optics Analysis - An Advanced 3D Trajectory Code

    International Nuclear Information System (INIS)

    Ives, R. Lawrence; Bui, Thuc; Vogler, William; Neilson, Jeff; Read, Mike; Shephard, Mark; Bauer, Andrew; Datta, Dibyendu; Beal, Mark

    2006-01-01

    Calabazas Creek Research, Inc. has completed initial development of an advanced, 3D program for modeling electron trajectories in electromagnetic fields. The code is being used to design complex guns and collectors. Beam Optics Analysis (BOA) is a fully relativistic, charged particle code using adaptive, finite element meshing. Geometrical input is imported from CAD programs generating ACIS-formatted files. Parametric data is inputted using an intuitive, graphical user interface (GUI), which also provides control of convergence, accuracy, and post processing. The program includes a magnetic field solver, and magnetic information can be imported from Maxwell 2D/3D and other programs. The program supports thermionic emission and injected beams. Secondary electron emission is also supported, including multiple generations. Work on field emission is in progress as well as implementation of computer optimization of both the geometry and operating parameters. The principle features of the program and its capabilities are presented

  6. Development Of The Computer Code For Comparative Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Purwadi, Mohammad Dhandhang

    2001-01-01

    The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation

  7. Three-field modeling with droplet entrainment and de-entrainment models for TRAC-M

    International Nuclear Information System (INIS)

    Lee, Sang Ik

    2005-02-01

    A three-field modeling has been developed and implemented to the basic one-dimensional components of TRAC-M/F90 (Modernized Transient Reactor Analysis Code, Fortran90) to improve the estimation of the behavior of droplet entrainment. the divide and conquer algorithmic technique is adjusted for the implementation after considering the verifications. The governing equations are composed of the conservation equations of each field with thermal-equilibrium assumption between the two liquid fields. As a result, three momentum equations, four mass equations including noncondensable gas mass equation and two energy equations are used. In the development of the momentum equation for an entrained droplet, two different kinds of approaches are tested: the simple force balance on a single droplet and the field-type equation based on the continuum assumption. The first showed unstable nature and the latter was chosen. All of the necessary empirical correlations were obtained from literatures, in particular, the physical models of COBRA-TF. The sensitivities of each empirical correlation were investigated where the effect of droplet diameter and droplet drag was found to be negligible. The advantages of the SETS numerical scheme in comparison with the semi-implicit nimerical scheme were addressed in the simulation time and the maximum allowed time-step size. Finally, the simulations of Collier and Hewitt's experiment were performed and it is concluded that the three-field model developed in the present study is reasonable. The predictions by the present code agree with the measured values within the RMS errors of 16.5% and 9.6% in the entrained liquid mass flow rates and the pressure gradients, respectively. Assessment on the physical models is performed. The wall and interfacial drag models were determined at first. After that, the assessments of entrainment model were performed. Wurtz's entrainment model, which was used in COBRA-TF and Kataoka and Ishii's model, were mainly

  8. The Analysis of SBWR Critical Power Bundle Using Cobrag Code

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2013-03-01

    Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

  9. TechTracS: NASA's commercial technology management system

    Science.gov (United States)

    Barquinero, Kevin; Cannon, Douglas

    1996-03-01

    The Commercial Technology Mission is a primary NASA mission, comparable in importance to those in aeronautics and space. This paper will discuss TechTracS, NASA Commercial Technology Management System that has been put into place in FY 1995 to implement this mission. This system is designed to identify and capture the NASA technologies which have commercial potential into an off-the-shelf database application, and then track the technologies' progress in realizing the commercial potential through collaborations with industry. The management system consists of four stages. The first is to develop an inventory database of the agency's entire technology portfolio and assess it for relevance to the commercial marketplace. Those technologies that are identified as having commercial potential will then be actively marketed to appropriate industries—this is the second stage. The third stage is when a NASA-industry partnership is entered into for the purposes of commercializing the technology. The final stage is to track the technology's success or failure in the marketplace. The collection of this information in TechTracS enables metrics evaluation and can accelerate the establishment on direct contacts between and NASA technologist and an industry technologist. This connection is the beginning of the technology commercialization process.

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

    Science.gov (United States)

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

    2010-10-01

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

  11. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

  12. A simple in-surge pressure analysis using the SPACE code

    International Nuclear Information System (INIS)

    Youn, Bum Soo; Kim, Yo Han; Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2010-01-01

    Currently, nuclear safety analysis codes used in Korea are developed by all the overseas. These codes are paying huge fee and permission must be obtained for use in the country. In addition, orders for nuclear power plants must ensure the safety analysis code for independent domestic technology. Therefore, Korea Electric Power Research Institute(KEPRI) is developing the domestic nuclear power safety analysis, SPACE(Safety and Performance Analysis Code for nuclear power plants). To determine the computational power of pressurizer model in development SPACE code, it was compared with existing commercial nuclear power safety analysis code, RETRAN

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

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Suzuki, Motoe

    2000-01-01

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

  14. Meanline Analysis of Turbines with Choked Flow in the Object-Oriented Turbomachinery Analysis Code

    Science.gov (United States)

    Hendricks, Eric S.

    2016-01-01

    The Object-Oriented Turbomachinery Analysis Code (OTAC) is a new meanline/streamline turbomachinery modeling tool being developed at NASA GRC. During the development process, a limitation of the code was discovered in relation to the analysis of choked flow in axial turbines. This paper describes the relevant physics for choked flow as well as the changes made to OTAC to enable analysis in this flow regime.

  15. RADTRAN 5 - A computer code for transportation risk analysis

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Kanipe, F.L.

    1993-01-01

    The RADTRAN 5 computer code has been developed to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI standard FORTRAN 77; the code contains significant advances in the methodology first pioneered with the LINK option of RADTRAN 4. A major application of the LINK methodology is route-specific analysis. Another application is comparisons of attributes along the same route segments. Nonradiological risk factors have been incorporated to allow users to estimate nonradiological fatalities and injuries that might occur during the transportation event(s) being analyzed. These fatalities include prompt accidental fatalities from mechanical causes. Values of these risk factors for the United States have been made available in the code as optional defaults. Several new health effects models have been published in the wake of the Hiroshima-Nagasaki dosimetry reassessment, and this has emphasized the need for flexibility in the RADTRAN approach to health-effects calculations. Therefore, the basic set of health-effects conversion equations in RADTRAN have been made user-definable. All parameter values can be changed by the user, but a complete set of default values are available for both the new International Commission on Radiation Protection model (ICRP Publication 60) and the recent model of the U.S. National Research Council's Committee on the Biological Effects of Radiation (BEIR V). The meteorological input data tables have been modified to permit optional entry of maximum downwind distances for each dose isopleth. The expected dose to an individual in each isodose area is also calculated and printed automatically. Examples are given that illustrate the power and flexibility of the RADTRAN 5 computer code. (J.P.N.)

  16. Available computer codes and data for radiation transport analysis

    International Nuclear Information System (INIS)

    Trubey, D.K.; Maskewitz, B.F.; Roussin, R.W.

    1975-01-01

    The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  17. Development validation and use of computer codes for inelastic analysis

    International Nuclear Information System (INIS)

    Jobson, D.A.

    1983-01-01

    A finite element scheme is a system which provides routines so carry out the operations which are common to all finite element programs. The list of items that can be provided as standard by the finite element scheme is surprisingly large and the list provided by the UNCLE finite element scheme is unusually comprehensive. This presentation covers the following: construction of the program, setting up a finite element mesh, generation of coordinates, incorporating boundary and load conditions. Program validation was done by creep calculations performed using CAUSE code. Program use is illustrated by calculating a typical inelastic analysis problem. This includes computer model of the PFR intermediate heat exchanger

  18. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

    International Nuclear Information System (INIS)

    Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

  19. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  20. Numerical analysis and nuclear standard code application to thermal fatigue

    International Nuclear Information System (INIS)

    Merola, M.

    1992-01-01

    The present work describes the Joint Research Centre Ispra contribution to the IAEA benchmark exercise 'Lifetime Behaviour of the First Wall of Fusion Machines'. The results of the numerical analysis of the reference thermal fatigue experiment are presented. Then a discussion on the numerical analysis of thermal stress is tackled, pointing out its particular aspects in view of their influence on the stress field evaluation. As far as the design-allowable number of cycles are concerned the American nuclear code ASME and the French code RCC-MR are applied and the reasons for the different results obtained are investigated. As regards a realistic fatigue lifetime evaluation, the main problems to be solved are brought out. This work, is intended as a preliminary basis for a discussion focusing on the main characteristics of the thermal fatigue problem from both a numerical and a lifetime assessment point of view. In fact the present margin of discretion left to the analyst may cause undue discrepancies in the results obtained. A sensitivity analysis of the main parameters involved is desirable and more precise design procedures should be stated

  1. Analysis of Non-binary Hybrid LDPC Codes

    OpenAIRE

    Sassatelli, Lucile; Declercq, David

    2008-01-01

    In this paper, we analyse asymptotically a new class of LDPC codes called Non-binary Hybrid LDPC codes, which has been recently introduced. We use density evolution techniques to derive a stability condition for hybrid LDPC codes, and prove their threshold behavior. We study this stability condition to conclude on asymptotic advantages of hybrid LDPC codes compared to their non-hybrid counterparts.

  2. Systemization of burnup sensitivity analysis code (2) (Contract research)

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  3. Comparative analysis of design codes for timber bridges in Canada, the United States, and Europe

    Science.gov (United States)

    James Wacker; James (Scott) Groenier

    2010-01-01

    The United States recently completed its transition from the allowable stress design code to the load and resistance factor design (LRFD) reliability-based code for the design of most highway bridges. For an international perspective on the LRFD-based bridge codes, a comparative analysis is presented: a study addressed national codes of the United States, Canada, and...

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

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.

    2001-01-01

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

  5. Investigation of alpha experiment by severe accident analysis code SAMPSON

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi; Naitoh, Masanori

    2006-01-01

    The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al 2 O 3 ) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the core to mimic the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined, where the results show its necessity in order to correctly reproduce the experimental trends. (author)

  6. Optimization and Validation of the Developed Uranium Isotopic Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.

  7. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  9. Analysis results from the Los Alamos 2D/3D program

    International Nuclear Information System (INIS)

    Boyack, B.E.; Cappiello, M.W.; Harmony, S.C.; Shire, P.R.; Siebe, D.A.

    1987-01-01

    Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test-facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multi-dimensional, nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During fiscal year 1987, Los Alamos conducted analytical assessment activities using data from the Slab Core Test Facility, The Cylindrical Core Test Facility, and the Upper Plenum Test Facility. Finally, Los Alamos continued work to provide TRAC improvements. In this paper, Los Alamos activities during fiscal year 1987 will be summarized; several significant accomplishments will be described in more detail to illustrate the work activities at Los Alamos

  10. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (vectorization). Progress report fiscal 1997

    International Nuclear Information System (INIS)

    Kawasaki, Nobuo; Ogasawara, Shinobu; Adachi, Masaaki; Kume, Etsuo; Ishizuki, Shigeru; Tanabe, Hidenobu; Nemoto, Toshiyuki; Kawai, Wataru; Watanabe, Hideo

    1999-05-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system and/or the AP3000 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 14 codes in fiscal 1997. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the vectorization. In this vectorization part, the vectorization of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations, statistical decay code SD and three-dimensional thermal analysis code for in-core test section (T2) of HENDEL SSPHEAT are described. In the parallelization part, the parallelization of cylindrical direct numerical simulation code CYLDNS44N, worldwide version of system for prediction of environmental emergency dose information code WSPEEDI, extension of quantum molecular dynamics code EQMD and three-dimensional non-steady compressible fluid dynamics code STREAM are described. In the porting part, the porting of transient reactor analysis code TRAC-BF1 and Monte Carlo radiation transport code MCNP4A on the AP3000 are described. In addition, a modification of program libraries for command-driven interactive data analysis plotting program IPLOT is described. (author)

  11. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (porting). Progress report fiscal 1997

    International Nuclear Information System (INIS)

    Ishizuki, Shigeru; Nemoto, Toshiyuki; Kawai, Wataru; Watanabe, Hideo; Tanabe, Hidenobu; Kawasaki, Nobuo; Adachi, Masaaki; Ogasawara, Shinobu; Kume, Etsuo

    1999-05-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system and/or the AP3000 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 14 codes in fiscal 1997. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the porting. In this porting part, the porting of transient reactor analysis code TRAC-BF1 and Monte Carlo radiation transport code MCNP4A on the AP3000 are described. In addition, a modification of program libraries for command-driven interactive data analysis plotting program IPLOT is described. In the vectorization part, the vectorization of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations, statistical decay code SD and three-dimensional thermal analysis code for in-core test section (T2) of HENDEL SSPHEAT are described. In the parallelization part, the parallelization of cylindrical direct numerical simulation code CYLDNS44N, worldwide version of system for prediction of environmental emergency dose information code WSPEEDI, extension of quantum molecular dynamics code EQMD and three-dimensional non-steady compressible fluid dynamics code STREAM are described. (author)

  12. Interface design of VSOP'94 computer code for safety analysis

    International Nuclear Information System (INIS)

    Natsir, Khairina; Andiwijayakusuma, D.; Wahanani, Nursinta Adi; Yazid, Putranto Ilham

    2014-01-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects

  13. Interface design of VSOP'94 computer code for safety analysis

    Science.gov (United States)

    Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

    2014-09-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

  14. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  15. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    International Nuclear Information System (INIS)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

    2014-01-01

    The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

  16. Corporate governance codes and their contents : An analysis of Eastern European codes

    NARCIS (Netherlands)

    Hermes, Niels; Postma, Theo J. B. M.; Zivkov, Orestis

    2007-01-01

    Existing literature suggests that the contents of corporate governance codes are similar due to external forces, such as increased integration of countries in the global economy, the increased role of foreign institutional investors and recommendations on corporate governance practices of

  17. COMTA - a computer code for fuel mechanical and thermal analysis

    International Nuclear Information System (INIS)

    Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.

    1979-01-01

    COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)

  18. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  19. Analysis of visual coding variables on CRT generated displays

    International Nuclear Information System (INIS)

    Blackman, H.S.; Gilmore, W.E.

    1985-01-01

    Cathode ray tube generated safety parameter display systems in a nuclear power plant control room situation have been found to be improved in effectiveness when color coding is employed. Research has indicated strong support for graphic coding techniques particularly in redundant coding schemes. In addition, findings on pictographs, as applied in coding schemes, indicate the need for careful application and for further research in the development of a standardized set of symbols

  20. Fuel analysis code FAIR and its high burnup modelling capabilities

    International Nuclear Information System (INIS)

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

    1995-01-01

    A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

  1. Ethical Code Effectiveness in Football Clubs: A Longitudinal Analysis

    OpenAIRE

    Constandt, Bram; De Waegeneer, Els; Willem, Annick

    2017-01-01

    As football (soccer) clubs are facing different ethical challenges, many clubs are turning to ethical codes to counteract unethical behaviour. However, both in- and outside the sport field, uncertainty remains about the effectiveness of these ethical codes. For the first time, a longitudinal study design was adopted to evaluate code effectiveness. Specifically, a sample of non-professional football clubs formed the subject of our inquiry. Ethical code effectiveness was...

  2. An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

    International Nuclear Information System (INIS)

    Min Lee

    1994-01-01

    Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data

  3. Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

    International Nuclear Information System (INIS)

    Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao

    2000-01-01

    Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

  4. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  5. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  6. DYNAVAC: a transient-vacuum-network analysis code

    International Nuclear Information System (INIS)

    Deis, G.A.

    1980-01-01

    This report discusses the structure and use of the program DYNAVAC, a new transient-vacuum-network analysis code implemented on the NMFECC CDC-7600 computer. DYNAVAC solves for the transient pressures in a network of up to twenty lumped volumes, interconnected in any configuration by specified conductances. Each volume can have an internal gas source, a pumping speed, and any initial pressure. The gas-source rates can vary with time in any piecewise-linear manner, and up to twenty different time variations can be included in a single problem. In addition, the pumping speed in each volume can vary with the total gas pumped in the volume, thus simulating the saturation of surface pumping. This report is intended to be both a general description and a user's manual for DYNAVAC

  7. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  8. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  9. Behavior Analysis Usage with Behavior Tures Adoption for Malicious Code Detection on JAVASCRIPT Scenarios Example

    Directory of Open Access Journals (Sweden)

    Y. M. Tumanov

    2010-03-01

    Full Text Available The article offers the method of malicious JavaScript code detection, based on behavior analysis. Conceptions of program behavior, program state, an algorithm of malicious code detection are described.

  10. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

  11. Turbine Airfoil Optimization Using Quasi-3D Analysis Codes

    Directory of Open Access Journals (Sweden)

    Sanjay Goel

    2009-01-01

    Full Text Available A new approach to optimize the geometry of a turbine airfoil by simultaneously designing multiple 2D sections of the airfoil is presented in this paper. The complexity of 3D geometry modeling is circumvented by generating multiple 2D airfoil sections and constraining their geometry in the radial direction using first- and second-order polynomials that ensure smoothness in the radial direction. The flow fields of candidate geometries obtained during optimization are evaluated using a quasi-3D, inviscid, CFD analysis code. An inviscid flow solver is used to reduce the execution time of the analysis. Multiple evaluation criteria based on the Mach number profile obtained from the analysis of each airfoil cross-section are used for computing a quality metric. A key contribution of the paper is the development of metrics that emulate the perception of the human designer in visually evaluating the Mach Number distribution. A mathematical representation of the evaluation criteria coupled with a parametric geometry generator enables the use of formal optimization techniques in the design. The proposed approach is implemented in the optimal design of a low-pressure turbine nozzle.

  12. Code of Practice for Protection Against Ionizing Radiation Emitted from X-ray Analysis Equipment (1984)

    International Nuclear Information System (INIS)

    1984-01-01

    Appropriate working rules, safety features and monitoring requirements for general X-ray analysis units or equipment are laid down in this Code which is intended for users of such equipment. The Code advises that establishments draw up their own working procedures based on appropriate legislation and on the recommendations contained in this Code. The Code also describes the requirements for X-ray analysis equipment necessary to ensure safety. (NEA) [fr

  13. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  14. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  15. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  16. Uncertainty and sensitivity analysis using probabilistic system assessment code. 1

    International Nuclear Information System (INIS)

    Honma, Toshimitsu; Sasahara, Takashi.

    1993-10-01

    This report presents the results obtained when applying the probabilistic system assessment code under development to the PSACOIN Level 0 intercomparison exercise organized by the Probabilistic System Assessment Code User Group in the Nuclear Energy Agency (NEA) of OECD. This exercise is one of a series designed to compare and verify probabilistic codes in the performance assessment of geological radioactive waste disposal facilities. The computations were performed using the Monte Carlo sampling code PREP and post-processor code USAMO. The submodels in the waste disposal system were described and coded with the specification of the exercise. Besides the results required for the exercise, further additional uncertainty and sensitivity analyses were performed and the details of these are also included. (author)

  17. Transient analysis of ABWR reactor using a best estimate code

    International Nuclear Information System (INIS)

    Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.

    2004-01-01

    Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)

  18. Analysis of fuel-coolant interaction with VAPEX code

    International Nuclear Information System (INIS)

    Melikhov, O.I.; Melikhov, V.I.; Sokolin, A.V.; Yakush, S.E.

    2004-01-01

    The analysis of the FARO L-33 test has been carried out with the VAPEX code in which a submodel for hydrogen release and transport was implemented. The FARO test was aimed at studying the premixing and quenching processes for large (about 100 kg) masses of corium. The specific features of the FARO L-33 test are: high subcooling (124 K), low pressure (4.1 bar), presence of non-condensable gas (argon) and triggered vapor explosion when melt reached the bottom of the vessel. A numerical simulation of FARO L-33 test was carried out using 2-D nodalization. The fragmentation model is based on the Saito correlation. The model for hydrogen release assumes direct proportionality between the total hydrogen mass release rate and the total fragmentation rate of the melt jet. The proportionality constant was taken from the experimental estimates for test conditions. Calculation of the premixing stage gave some delay in the pressure growth, which is most probably connected with inadequacy of the fragmentation model at the initial stage of melt jet-water interaction. The calculated pressurization rate, however, agrees reasonably with the measured one. Modeling of vapor explosion, which occurred in the test, yielded reasonable correlation with the test data when hydrogen formation was taken into account. Thus, VAPEX analysis of the FARO L-33 test has shown reasonable agreement between the experimental and calculated data. (author)

  19. Analysis of Memory Codes and Cumulative Rehearsal in Observational Learning

    Science.gov (United States)

    Bandura, Albert; And Others

    1974-01-01

    The present study examined the influence of memory codes varying in meaningfulness and retrievability and cumulative rehearsal on retention of observationally learned responses over increasing temporal intervals. (Editor)

  20. TRANSURANUS: A fuel rod analysis code ready for use

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Van de Laar, J [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Ott, C [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (1{sub 1/2}-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs.

  1. Development of severe accident analysis code - Development of a finite element code for lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-08-01

    The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  3. Code development for eigenvalue total sensitivity analysis and total uncertainty analysis

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Zu, Tiejun; Shen, Wei

    2015-01-01

    Highlights: • We develop a new code for total sensitivity and uncertainty analysis. • The implicit effects of cross sections can be considered. • The results of our code agree well with TSUNAMI-1D. • Detailed analysis for origins of implicit effects is performed. - Abstract: The uncertainties of multigroup cross sections notably impact eigenvalue of neutron-transport equation. We report on a total sensitivity analysis and total uncertainty analysis code named UNICORN that has been developed by applying the direct numerical perturbation method and statistical sampling method. In order to consider the contributions of various basic cross sections and the implicit effects which are indirect results of multigroup cross sections through resonance self-shielding calculation, an improved multigroup cross-section perturbation model is developed. The DRAGON 4.0 code, with application of WIMSD-4 format library, is used by UNICORN to carry out the resonance self-shielding and neutron-transport calculations. In addition, the bootstrap technique has been applied to the statistical sampling method in UNICORN to obtain much steadier and more reliable uncertainty results. The UNICORN code has been verified against TSUNAMI-1D by analyzing the case of TMI-1 pin-cell. The numerical results show that the total uncertainty of eigenvalue caused by cross sections can reach up to be about 0.72%. Therefore the contributions of the basic cross sections and their implicit effects are not negligible

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  5. Description and validation of ANTEO, an optimised PC code the thermalhydraulic analysis of fuel bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1995-01-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of such a code was made possible by two facts: firstly, the increase, in the computing power of the desk machines; secondly, the fact that several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes. (author)

  6. ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1996-07-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes

  7. Performance Analysis of Optical Code Division Multiplex System

    Science.gov (United States)

    Kaur, Sandeep; Bhatia, Kamaljit Singh

    2013-12-01

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

  8. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    International Nuclear Information System (INIS)

    Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E.

    2002-01-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  9. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    Energy Technology Data Exchange (ETDEWEB)

    Valtonen, Keijo (Radiation and Nuclear Safety Authority, Finland); Hamalainen, Anitta (VTT Energy, Finland); Cunningham, Mitchel E.(BATTELLE (PACIFIC NW LAB))

    2002-05-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  10. Sample problem manual for benchmarking of cask analysis codes

    International Nuclear Information System (INIS)

    Glass, R.E.

    1988-02-01

    A series of problems have been defined to evaluate structural and thermal codes. These problems were designed to simulate the hypothetical accident conditions given in Title 10 of the Code of Federal Regulation, Part 71 (10CFR71) while retaining simple geometries. This produced a problem set that exercises the ability of the codes to model pertinent physical phenomena without requiring extensive use of computer resources. The solutions that are presented are consensus solutions based on computer analyses done by both national laboratories and industry in the United States, United Kingdom, France, Italy, Sweden, and Japan. The intent of this manual is to provide code users with a set of standard structural and thermal problems and solutions which can be used to evaluate individual codes. 19 refs., 19 figs., 14 tabs

  11. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    Belac, J.; Josek, R.; Klecka, L.; Stary, V.; Vocka, R.

    2005-01-01

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  12. MMA, A Computer Code for Multi-Model Analysis

    Science.gov (United States)

    Poeter, Eileen P.; Hill, Mary C.

    2007-01-01

    This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will

  13. The CAIN computer code for the generation of MABEL input data sets: a user's manual

    International Nuclear Information System (INIS)

    Tilley, D.R.

    1983-03-01

    CAIN is an interactive FORTRAN computer code designed to overcome the substantial effort involved in manually creating the thermal-hydraulics input data required by MABEL-2. CAIN achieves this by processing output from either of the whole-core codes, RELAP or TRAC, interpolating where necessary, and by scanning RELAP/TRAC output in order to generate additional information. This user's manual describes the actions required in order to create RELAP/TRAC data sets from magnetic tape, to create the other input data sets required by CAIN, and to operate the interactive command procedure for the execution of CAIN. In addition, the CAIN code is described in detail. This programme of work is part of the Nuclear Installations Inspectorate (NII)'s contribution to the United Kingdom Atomic Energy Authority's independent safety assessment of pressurized water reactors. (author)

  14. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  15. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  16. Safety analysis and code development for nuclear fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    We are estimating that the debris containing fuel are piled in the containment and the pressure vessel bottoms of Fukushima-Daiichi NPPs. A radioactive Xe concentration discharged in recriticality is being monitored by utilizing the gas management system set up in NPPs unit 1-3. For this reason, we can confirm the recriticality might not be broken out. However, the debris conditions distributed in the containment vessel and the pressure vessel bottoms are not clear. The internal and external surrounding changes will make recriticality become possible. According to TEPCO's roadmap, TEPCO will launch extracting task within 10 years. Even in the case that the fuel condition changes due to debris relocation and mixture, subcriticality must be secured. Criticality safety analysis with non-uniform effect should therefore be essential for the molten debris. For above reasons, we studies the optimum distributions with some parameters that have a large reactivity change were assessed with OPT-DANT code. Finally, the boron concentration was estimated in order to keep subcriticality. (author)

  17. Flow analysis of tubular fuel assembly using CFD code

    International Nuclear Information System (INIS)

    Park, J. H.; Park, C.; Chae, H. T.

    2004-01-01

    Based on the experiences of HANARO, a new research reactor is under conceptual design preparing for future needs of research reactor. Considering various aspects such as nuclear physics, thermal-hydraulics, mechanical structure and the applicability of HANARO technology, a tubular type fuel has been considered as that of a new research reactor. Tubular type fuel has several circular fuel layers, and each layer consists of 3 curved fuel plates arranged with constant small gap to build up cooling channels. In the thermal-hydraulic point, it is very important to maintain each channel flow velocity be equal as much as possible, because the small gaps between curved thin fuel plates independently forms separate coolant channels, which may cause a thermal-hydraulic problem in certain conditions. In this study, commercial CFD(Computational Fluid Dynamics) code, Fluent, has been used to investigate flow characteristics of tubular type fuel assembly. According to the computation results for the preliminary conceptual design, there is a serious lack of uniformity of average velocity on the each coolant channel. Some changes for initial conceptual design were done to improve the balance of velocity distribution, and analysis was done again, too. The results for the revised design showed that the uniformity of each channel velocity was improved significantly. The influence of outermost channel gap width on the velocity distribution was also examined

  18. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

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

    Directory of Open Access Journals (Sweden)

    O. Al Rasheed

    2014-06-01

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

  20. Feasibility study for objective oriented design of system thermal hydraulic analysis program

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu

    2008-01-01

    The system safety analysis code, such as RELAP5, TRAC, CATHARE etc. have been developed based on Fortran language during the past few decades. Refactoring of conventional codes has been also performed to improve code readability and maintenance. However the programming paradigm in software technology has been changed to use objects oriented programming (OOP), which is based on several techniques, including encapsulation, modularity, polymorphism, and inheritance. In this work, objective oriented program for system safety analysis code has been tried utilizing modernized C language. The analysis, design, implementation and verification steps for OOP system code development are described with some implementation examples. The system code SYSTF based on three-fluid thermal hydraulic solver has been developed by OOP design. The verifications of feasibility are performed with simple fundamental problems and plant models. (author)

  1. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

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

    National Research Council Canada - National Science Library

    Ong, Choon

    1998-01-01

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

  3. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    International Nuclear Information System (INIS)

    Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

    2015-01-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  4. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  5. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Liu Chun; Wang Chang; Cao Jun-Cheng

    2016-01-01

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

  7. Computer codes for beam dynamics analysis of cyclotronlike accelerators

    Science.gov (United States)

    Smirnov, V.

    2017-12-01

    Computer codes suitable for the study of beam dynamics in cyclotronlike (classical and isochronous cyclotrons, synchrocyclotrons, and fixed field alternating gradient) accelerators are reviewed. Computer modeling of cyclotron segments, such as the central zone, acceleration region, and extraction system is considered. The author does not claim to give a full and detailed description of the methods and algorithms used in the codes. Special attention is paid to the codes already proven and confirmed at the existing accelerating facilities. The description of the programs prepared in the worldwide known accelerator centers is provided. The basic features of the programs available to users and limitations of their applicability are described.

  8. Development of safety analysis codes for light water reactor

    International Nuclear Information System (INIS)

    Akimoto, Masayuki

    1985-01-01

    An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

  9. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  10. A comparison of LOCA analysis using SMOKIN and CERBERUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.

  11. Vehicle Codes and Standards: Overview and Gap Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Blake, C.; Buttner, W.; Rivkin, C.

    2010-02-01

    This report identifies gaps in vehicle codes and standards and recommends ways to fill the gaps, focusing on six alternative fuels: biodiesel, natural gas, electricity, ethanol, hydrogen, and propane.

  12. Exposure calculation code module for reactor core analysis: BURNER

    International Nuclear Information System (INIS)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

  13. Evolutionary analysis reveals regulatory and functional landscape of coding and non-coding RNA editing.

    Science.gov (United States)

    Zhang, Rui; Deng, Patricia; Jacobson, Dionna; Li, Jin Billy

    2017-02-01

    Adenosine-to-inosine RNA editing diversifies the transcriptome and promotes functional diversity, particularly in the brain. A plethora of editing sites has been recently identified; however, how they are selected and regulated and which are functionally important are largely unknown. Here we show the cis-regulation and stepwise selection of RNA editing during Drosophila evolution and pinpoint a large number of functional editing sites. We found that the establishment of editing and variation in editing levels across Drosophila species are largely explained and predicted by cis-regulatory elements. Furthermore, editing events that arose early in the species tree tend to be more highly edited in clusters and enriched in slowly-evolved neuronal genes, thus suggesting that the main role of RNA editing is for fine-tuning neurological functions. While nonsynonymous editing events have been long recognized as playing a functional role, in addition to nonsynonymous editing sites, a large fraction of 3'UTR editing sites is evolutionarily constrained, highly edited, and thus likely functional. We find that these 3'UTR editing events can alter mRNA stability and affect miRNA binding and thus highlight the functional roles of noncoding RNA editing. Our work, through evolutionary analyses of RNA editing in Drosophila, uncovers novel insights of RNA editing regulation as well as its functions in both coding and non-coding regions.

  14. Development of in-vessel source term analysis code, tracer

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

    Analyses of radionuclide transport in fuel failure accidents (generally referred to source terms) are considered to be important especially in the severe accident evaluation. The TRACER code has been developed to realistically predict the time dependent behavior of FPs and aerosols within the primary cooling system for wide range of fuel failure events. This paper presents the model description, results of validation study, the recent model advancement status of the code, and results of check out calculations under reactor conditions. (author)

  15. Demonstration study on shielding safety analysis code (8)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

    2001-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

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

    DEFF Research Database (Denmark)

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

    1998-01-01

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

  17. Sensitivity analysis of FRAPCON-1 computer code to some parameters

    International Nuclear Information System (INIS)

    Chia, C.T.; Silva, C.F. da.

    1987-05-01

    A sensibility study of the code FRAPCON-1 was done for the following inout data: number of axial nodes, number of time steps and the axial power shape. Their influence in the code response concerning to the fuel center line temperature, stored energy, internal gas pressure, clad hoop strain and gap width were analyzed. The number of axial nodes has little influence, but care must be taken in the choice of the power axial profile and the time step length. (Author) [pt

  18. Horizontal stratified flow model for the 1-D module of WCOBRA/TRAC-TF2: modeling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Frepoli, C.; Ohkawa, K., E-mail: liaoj@westinghouse.com [Westinghouse Electric Company LLC, LOCA Integrated Services I, Cranberry Twp, Pennsylvania (United States)

    2011-07-01

    For a two-phase flow in a horizontal pipe, the individual phases may separate by gravity. This horizontal stratification significantly impacts the interfacial drag, interfacial heat transfer and wall drag of the two phase flow. For a PWR small break LOCA, the horizontal stratification in cold legs is a highly important phenomenon during loop seal clearance, boiloff and recovery periods. The low interfacial drag in the stratified flow directly controls the time period for the loop clearance and the level of residual water in the loop seal. Horizontal stratification in hot legs also impacts the natural circulation stage of a small break LOCA. In addition, the offtake phenomenon and cold leg condensation phenomenon are also affected by the occurrence of horizontal stratification in the cold legs. In the 1-D module of the WCOBRA/TRAC-TF2 computer code, a horizontal stratification criterion was developed by combining the Taitel-Dukler model and the Wallis-Dobson model, which approximates the viscous Kelvin-Helmholtz neutral stability boundary. The objective of this paper is to present the horizontal stratification model implemented in the code and its assessment against relevant data. The adequacy of the horizontal stratification transition criterion is confirmed by examining the code-predicted flow regime in a horizontal pipe with the measured data in the flow regime map. The void fractions (or liquid level) for the horizontal stratified flow in cold leg or hot leg are predicted with a reasonable accuracy. (author)

  19. Code Blue Emergencies: A Team Task Analysis and Educational Initiative.

    Science.gov (United States)

    Price, James W; Applegarth, Oliver; Vu, Mark; Price, John R

    2012-01-01

    The objective of this study was to identify factors that have a positive or negative influence on resuscitation team performance during emergencies in the operating room (OR) and post-operative recovery unit (PAR) at a major Canadian teaching hospital. This information was then used to implement a team training program for code blue emergencies. In 2009/10, all OR and PAR nurses and 19 anesthesiologists at Vancouver General Hospital (VGH) were invited to complete an anonymous, 10 minute written questionnaire regarding their code blue experience. Survey questions were devised by 10 recovery room and operation room nurses as well as 5 anesthesiologists representing 4 different hospitals in British Columbia. Three iterations of the survey were reviewed by a pilot group of nurses and anesthesiologists and their feedback was integrated into the final version of the survey. Both nursing staff (n = 49) and anesthesiologists (n = 19) supported code blue training and believed that team training would improve patient outcome. Nurses noted that it was often difficult to identify the leader of the resuscitation team. Both nursing staff and anesthesiologists strongly agreed that too many people attending the code blue with no assigned role hindered team performance. Identifiable leadership and clear communication of roles were identified as keys to resuscitation team functioning. Decreasing the number of people attending code blue emergencies with no specific role, increased access to mock code blue training, and debriefing after crises were all identified as areas requiring improvement. Initial team training exercises have been well received by staff.

  20. Business Ethics: International Analysis of Codes of Ethics and Conduct

    Directory of Open Access Journals (Sweden)

    Josmar Andrade

    2017-03-01

    Full Text Available Codes of ethics and code of conduct formalize an ideal of expected behavior patterns to managers and employees of organizations, providing standards and orientation that states companies interactions with the community, through products /services, sales force, marketing communications, investments, and relationships with other stakeholders, influencing company reputation and overall Marketing performance. The objective of this study is to analyze the differences in codes of ethics of the largest companies based in Brazil and in Portugal, given their cultural and linguistic similarities. Findings show that the use of codes of ethics are more common in Brazil than in Portugal and that codes of ethics are substantially more extensive and cover a larger number of categories in Brazilian companies, reflecting the organizations’ mission and perception of stakeholders concerns and priorities. We conclude that ethical issues severely impact company reputation and, in a comprehensive sense, overall Marketing performance. Marketing professionals should be systematically aware of how company core values are transmitted to different audiences, including the use of code of ethics to communicate both with internal and external publics. 0 0 1 171 966 CASA DOS ANDRADES 23 14 1123 14.0 96 800x600 Normal 0 false false false EN-US JA X-NONE  

  1. Promoter Analysis Reveals Globally Differential Regulation of Human Long Non-Coding RNA and Protein-Coding Genes

    KAUST Repository

    Alam, Tanvir

    2014-10-02

    Transcriptional regulation of protein-coding genes is increasingly well-understood on a global scale, yet no comparable information exists for long non-coding RNA (lncRNA) genes, which were recently recognized to be as numerous as protein-coding genes in mammalian genomes. We performed a genome-wide comparative analysis of the promoters of human lncRNA and protein-coding genes, finding global differences in specific genetic and epigenetic features relevant to transcriptional regulation. These two groups of genes are hence subject to separate transcriptional regulatory programs, including distinct transcription factor (TF) proteins that significantly favor lncRNA, rather than coding-gene, promoters. We report a specific signature of promoter-proximal transcriptional regulation of lncRNA genes, including several distinct transcription factor binding sites (TFBS). Experimental DNase I hypersensitive site profiles are consistent with active configurations of these lncRNA TFBS sets in diverse human cell types. TFBS ChIP-seq datasets confirm the binding events that we predicted using computational approaches for a subset of factors. For several TFs known to be directly regulated by lncRNAs, we find that their putative TFBSs are enriched at lncRNA promoters, suggesting that the TFs and the lncRNAs may participate in a bidirectional feedback loop regulatory network. Accordingly, cells may be able to modulate lncRNA expression levels independently of mRNA levels via distinct regulatory pathways. Our results also raise the possibility that, given the historical reliance on protein-coding gene catalogs to define the chromatin states of active promoters, a revision of these chromatin signature profiles to incorporate expressed lncRNA genes is warranted in the future.

  2. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  3. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-01

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error

  4. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  5. Analysis of the KUCA MEU experiments using the ANL code system

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.

  6. Analysis of the KUCA MEU experiments using the ANL code system

    International Nuclear Information System (INIS)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis

  7. Convolutional Sparse Coding for Static and Dynamic Images Analysis

    Directory of Open Access Journals (Sweden)

    B. A. Knyazev

    2014-01-01

    Full Text Available The objective of this work is to improve performance of static and dynamic objects recognition. For this purpose a new image representation model and a transformation algorithm are proposed. It is examined and illustrated that limitations of previous methods make it difficult to achieve this objective. Static images, specifically handwritten digits of the widely used MNIST dataset, are the primary focus of this work. Nevertheless, preliminary qualitative results of image sequences analysis based on the suggested model are presented.A general analytical form of the Gabor function, often employed to generate filters, is described and discussed. In this research, this description is required for computing parameters of responses returned by our algorithm. The recursive convolution operator is introduced, which allows extracting free shape features of visual objects. The developed parametric representation model is compared with sparse coding based on energy function minimization.In the experimental part of this work, errors of estimating the parameters of responses are determined. Also, parameters statistics and their correlation coefficients for more than 106 responses extracted from the MNIST dataset are calculated. It is demonstrated that these data correspond well with previous research studies on Gabor filters as well as with works on visual cortex primary cells of mammals, in which similar responses were observed. A comparative test of the developed model with three other approaches is conducted; speed and accuracy scores of handwritten digits classification are presented. A support vector machine with a linear or radial basic function is used for classification of images and their representations while principal component analysis is used in some cases to prepare data beforehand. High accuracy is not attained due to the specific difficulties of combining our model with a support vector machine (a 3.99% error rate. However, another method is

  8. Analysis of results of AZTRAN and AZKIND codes for a BWR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L.; Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M.

    2016-09-01

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  9. Non-Binary Protograph-Based LDPC Codes: Analysis,Enumerators and Designs

    OpenAIRE

    Sun, Yizeng

    2013-01-01

    Non-binary LDPC codes can outperform binary LDPC codes using sum-product algorithm with higher computation complexity. Non-binary LDPC codes based on protographs have the advantage of simple hardware architecture. In the first part of this thesis, we will use EXIT chart analysis to compute the thresholds of different protographs over GF(q). Based on threshold computation, some non-binary protograph-based LDPC codes are designed and their frame error rates are compared with binary LDPC codes. ...

  10. Development of statistical analysis code for meteorological data (W-View)

    International Nuclear Information System (INIS)

    Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori

    2003-03-01

    A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

  11. PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping

    International Nuclear Information System (INIS)

    Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.

    1975-01-01

    To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report

  12. Analysis of the sodium concrete interactions with the NABE code

    International Nuclear Information System (INIS)

    Soule, N.

    1989-01-01

    Experimental studies have been performed in France to investigate sodium-concrete interactions: thermal decomposition of concrete, specific chemical reactions, experimentation in liquid and vapour phase, sodium-concrete interaction without liner protection. Simultaneously computer codes have been developed in order to study the response of the containment building of a liquid metal fast breeder reactor to a sodium pool fire worsened by a sodium-concrete interaction: the NABE code. This code takes into account: a) sodium combustion; b) thermal decomposition of concrete with associated chemical reactions: (liquid sodium-vapour water reaction, liquid sodium-carbon dioxide reaction, liquid sodium-solid compounds of concrete, hydrogen combustion); c) chemical reactions in vapour phase; d) decay heat; e) gas aerosol inlets/outlets; f) aerosol behaviour (sedimentation, diffusion, leak); g) thermal exchanges. An example of a situation, typical of assessment of beyond design basis situations in LMFBR, is given. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-25

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

  14. Design and Analysis of LT Codes with Decreasing Ripple Size

    DEFF Research Database (Denmark)

    Sørensen, Jesper Hemming; Popovski, Petar; Østergaard, Jan

    2012-01-01

    In this paper we propose a new design of LT codes, which decreases the amount of necessary overhead in comparison to existing designs. The design focuses on a parameter of the LT decoding process called the ripple size. This parameter was also a key element in the design proposed in the original...... work by Luby. Specifically, Luby argued that an LT code should provide a constant ripple size during decoding. In this work we show that the ripple size should decrease during decoding, in order to reduce the necessary overhead. Initially we motivate this claim by analytical results related...... to the redundancy within an LT code. We then propose a new design procedure, which can provide any desired achievable decreasing ripple size. The new design procedure is evaluated and compared to the current state of the art through simulations. This reveals a significant increase in performance with respect...

  15. Code Blue Emergencies: A Team Task Analysis and Educational Initiative

    Directory of Open Access Journals (Sweden)

    James W. Price

    2012-04-01

    Full Text Available Introduction: The objective of this study was to identify factors that have a positive or negative influence on resuscitation team performance during emergencies in the operating room (OR and post-operative recovery unit (PAR at a major Canadian teaching hospital. This information was then used to implement a team training program for code blue emergencies. Methods: In 2009/10, all OR and PAR nurses and 19 anesthesiologists at Vancouver General Hospital (VGH were invited to complete an anonymous, 10 minute written questionnaire regarding their code blue experience. Survey questions were devised by 10 recovery room and operation room nurses as well as 5 anesthesiologists representing 4 different hospitals in British Columbia. Three iterations of the survey were reviewed by a pilot group of nurses and anesthesiologists and their feedback was integrated into the final version of the survey. Results: Both nursing staff (n = 49 and anesthesiologists (n = 19 supported code blue training and believed that team training would improve patient outcome. Nurses noted that it was often difficult to identify the leader of the resuscitation team. Both nursing staff and anesthesiologists strongly agreed that too many people attending the code blue with no assigned role hindered team performance. Conclusion: Identifiable leadership and clear communication of roles were identified as keys to resuscitation team functioning. Decreasing the number of people attending code blue emergencies with no specific role, increased access to mock code blue training, and debriefing after crises were all identified as areas requiring improvement. Initial team training exercises have been well received by staff.

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

    NARCIS (Netherlands)

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

    2014-01-01

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

  17. Improvements to the COBRA-TF (EPRI) computer code for steam generator analysis. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Barnhart, J.S.; Koontz, A.S.

    1980-09-01

    The COBRA-TF (EPRI) code has been improved and extended for pressurized water reactor steam generator analysis. New features and models have been added in the areas of subcooled boiling and heat transfer, turbulence, numerics, and global steam generator modeling. The code's new capabilities are qualified against selected experimental data and demonstrated for typical global and microscale steam generator analysis

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. General data analysis code for TDCR liquid scintillation counting

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-15

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

  20. General data analysis code for TDCR liquid scintillation counting

    International Nuclear Information System (INIS)

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

    2008-01-01

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