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Sample records for systems code development

  1. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  2. HELIAS module development for systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

    2015-02-15

    In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

  3. Recent developments in the Los Alamos radiation transport code system

    International Nuclear Information System (INIS)

    Forster, R.A.; Parsons, K.

    1997-01-01

    A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results

  4. 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)

  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. 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.

  7. 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

  8. 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

  9. 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)

  10. Development of FBR integrity system code. Basic concept

    International Nuclear Information System (INIS)

    Asayama, Tai

    2001-05-01

    For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

  11. An Expert System for the Development of Efficient Parallel Code

    Science.gov (United States)

    Jost, Gabriele; Chun, Robert; Jin, Hao-Qiang; Labarta, Jesus; Gimenez, Judit

    2004-01-01

    We have built the prototype of an expert system to assist the user in the development of efficient parallel code. The system was integrated into the parallel programming environment that is currently being developed at NASA Ames. The expert system interfaces to tools for automatic parallelization and performance analysis. It uses static program structure information and performance data in order to automatically determine causes of poor performance and to make suggestions for improvements. In this paper we give an overview of our programming environment, describe the prototype implementation of our expert system, and demonstrate its usefulness with several case studies.

  12. Development of particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)

  13. Development of a domestically-made system code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from the Fukushima-Daiichi NPP accidents, a new safety standard based on state-of-the-art findings has been established by the Japanese Nuclear Regulation Authority (NRA) and will soon come into force in Japan. In order to ensure a precise response to this movement from a technological point of view, it should be required for safety regulation to develop a new system code with much smaller uncertainty and reinforced simulation capability even in application to beyond-DBAs (BDBAs), as well as with the capability of close coupling to a newly developing severe accident code. Accordingly, development of a new domestically-made system code that incorporates 3-dimensional and 3 or more fluid thermal-hydraulics in tandem with a 3-dimensional neutronics has been started in 2012. In 2012, two branches of development activities, the development of 'main body' and advanced features have been started in parallel for development efficiency. The main body has been started from scratch and the following activities have therefore been performed: 1) development and determination of key principles and methodologies to realize a flexible, extensible and robust platform, 2) determination of requirements definition, 3) start of basic program design and coding and 4) start of a development of prototypical GUI-based pre-post processor. As for the advanced features, the following activities have been performed: 1) development of Phenomena Identification and Ranking Tables (PIRTs) and model capability matrix from normal operations to BDBAs in order to address requirements definition for advanced modeling, 2) development of detailed action plan for modification of field equations, numerical schemes and solvers and 3) start of the program development of field equations with an interfacial area concentration transport equation, a robust solver for condensation induced water hammer phenomena and a versatile Newton-Raphson solver. (author)

  14. Development of GUI systems for the MIDAS code

    International Nuclear Information System (INIS)

    Kim, K.R.; Park, S.H.; Kim, D.H.

    2004-01-01

    MIDAS is being developed at KAERI based on MELCOR as an integrated severe accident analysis code with existing model modification and new model addition. MIDAS was restructured to avoid the pointer based variable referencing style of MELCOR, and enhanced the memory effectiveness using the dynamic allocation method of Fortran 90. This paper describes recent activities of developing the GUI environments for MIDAS code at KAERI. Up to now, we have developed the four PC-based subsystems, which are IEDIT, IPLOT, SATS and HyperKAMG. IEDIT is an input management system that can read MELCOR input files and display its information in the Window panels. Users can modify each item in the panel and the input file will be modified according to that changes. IPLOT is a simple plotting system that can draw MIDAS plot variables trend graphs. SATS is developed as a severe accident training simulator that can display nuclear plant behavior graphically. Moreover SATS provides several controllable pumps and valves which appeared in the severe accidence. Together with SATS and the online severe accident guidance HyperKAMG, combined properly, severe accident mitigation scenarios could be presented graphically and dramatically without any change of MELCOR inputs. GUI development as a part of a severe accident management program package, MIDAS. (author)

  15. 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)

  16. Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

    2007-01-01

    EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

  17. Development of System Based Code: Case Study of Life-Cycle Margin Evaluation

    International Nuclear Information System (INIS)

    Tai Asayama; Masaki Morishita; Masanori Tashimo

    2006-01-01

    For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code. (authors)

  18. 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)

  19. Development and verification of a coupled code system RETRAN-MASTER-TORC

    International Nuclear Information System (INIS)

    Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

    2004-01-01

    Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

  20. Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan.

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  1. Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  2. 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

  3. 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

  4. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2015-01-01

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

  5. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany)

    2015-04-15

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

  6. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  7. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

  8. Development of a system of computer codes for severe accident analyses and its applications

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

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

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

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

  11. Code development of the national hemovigilance system and expansion strategies for hospital blood banks

    Directory of Open Access Journals (Sweden)

    Kim Jeongeun

    2012-01-01

    Full Text Available Objectives : The aims of this study were to develop reportable event codes that are applicable to the national hemovigilance systems for hospital blood banks, and to present expansion strategies for the blood banks. Materials and Methods : The data were obtained from a literature review and expert consultation, followed by adding to and revising the established hemovigilance code system and guidelines to develop reportable event codes for hospital blood banks. The Medical Error Reporting System-Transfusion Medicine developed in the US and other codes of reportable events were added to the Korean version of the Biologic Products Deviation Report (BPDR developed by the Korean Red Cross Blood Safety Administration, then using these codes, mapping work was conducted. We deduced outcomes suitable for practice, referred to the results of the advisory councils, and conducted a survey with experts and blood banks practitioners. Results : We developed reportable event codes that were applicable to hospital blood banks and could cover blood safety - from blood product safety to blood transfusion safety - and also presented expansion strategies for hospital blood banks. Conclusion : It was necessary to add 10 major categories to the blood transfusion safety stage and 97 reportable event codes to the blood safety stage. Contextualized solutions were presented on 9 categories of expansion strategies of hemovigilance system for the hospital blood banks.

  12. Development of the computer code system for the analyses of PWR core

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

    This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

  13. 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

  14. 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

  15. The development of efficient coding for an electronic mail system

    Science.gov (United States)

    Rice, R. F.

    1983-01-01

    Techniques for efficiently representing scanned electronic documents were investigated. Major results include the definition and preliminary performance results of a Universal System for Efficient Electronic Mail (USEEM), offering a potential order of magnitude improvement over standard facsimile techniques for representing textual material.

  16. Development of environmental dose assessment system (EDAS) code of PC version

    CERN Document Server

    Taki, M; Kobayashi, H; Yamaguchi, T

    2003-01-01

    A computer code (EDAS) was developed to assess the public dose for the safety assessment to get the license of nuclear reactor operation. This code system is used for the safety analysis of public around the nuclear reactor in normal operation and severe accident. This code was revised and composed for personal computer user according to the Nuclear Safety Guidelines reflected the ICRP1990 recommendation. These guidelines are revised by Nuclear Safety Commission on March, 2001, which are 'Weather analysis guideline for the safety assessment of nuclear power reactor', 'Public dose around the facility assessment guideline corresponding to the objective value for nuclear power light water reactor' and 'Public dose assessment guideline for safety review of nuclear power light water reactor'. This code has been already opened for public user by JAERI, and English version code and user manual are also prepared. This English version code is helpful for international cooperation concerning the nuclear safety assessme...

  17. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  18. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  19. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  20. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  1. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  2. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    International Nuclear Information System (INIS)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

    2007-03-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow

  3. Development of new two-dimensional spectral/spatial code based on dynamic cyclic shift code for OCDMA system

    Science.gov (United States)

    Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria

    2017-07-01

    In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.

  4. Development of the next generation code system as an engineering modeling language. (2). Study with prototyping

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto

    2003-04-01

    In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomena to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. Aiming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1) Multi-language (SoftWIRE.NET, Visual Basic.NET and Fortran) (2) Fortran 90 and (3) Python to make a prototype of the next generation code system. As this result, the followings were confirmed. (1) It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using Visual Basic.NET. (2) The maintainability of the existing code written by Fortran 77 can be improved by using the new features of Fortran 90. (3) The toolbox-type code system can be built by using Python. (author)

  5. CODE SWITCHING AND THE DEVELOPMENT OF LINGUISTIC SYSTEM OF SIMULTANEOUS BILINGUAL CHILDREN

    Directory of Open Access Journals (Sweden)

    Leni Amelia Suek

    2017-11-01

    Full Text Available Code switching and code mixing are the phenomena commonly seen done by a bilingual. This behavior is influenced by several aspects such as the linguistic system, sociolinguistics, pragmatics, and language competence of the bilingual. If children are able to distinguish two different languages since early age, they will be considered simultaneous bilinguals. They show that they develop multiple, rather than single, linguistic systems. However, it was understood that code switching and code mixing were due to the failure in using proper words, language features, and sociolinguistic competence. Yet, recent studies have shown that bilingual children are able to use both languages proficiently with no signs of confusion or failure in language use. This ability also does not hinder their cognitive development.

  6. Strategies for developing subchannel capability in an advanced system thermalhydraulic code: a literature review

    International Nuclear Information System (INIS)

    Cheng, J.; Rao, Y.F.

    2015-01-01

    In the framework of developing next generation safety analysis tools, Canadian Nuclear Laboratories (CNL) has planned to incorporate subchannel analysis capability into its advanced system thermalhydraulic code CATHENA 4. This paper provides a literature review and an assessment of current subchannel codes. It also evaluates three code-development methods: (i) static coupling of CATHENA 4 with the subchannel code ASSERT-PV, (ii) dynamic coupling of the two codes, and (iii) fully implicit implementation for a new, standalone CATHENA 4 version with subchannel capability. Results of the review and assessment suggest that the current ASSERT-PV modules can be used as the base for the fully implicit implementation of subchannel capability in CATHENA 4, and that this option may be the most cost-effective in the long run, resulting in savings in user application and maintenance costs. In addition, improved versatility of the tool could be accomplished by the addition of new features that could be added as part of its development. The new features would improve the capabilities of the existing subchannel code in handling low, reverse, and stagnant flows often encountered in system thermalhydraulic analysis. Therefore, the method of fully implicit implementation is preliminarily recommended for further exploration. A feasibility study will be performed in an attempt to extend the present work into a preliminary development plan. (author)

  7. Recent development for the ITS code system: Parallel processing and visualization

    International Nuclear Information System (INIS)

    Fan, W.C.; Turner, C.D.; Halbleib, J.A. Sr.; Kensek, R.P.

    1996-01-01

    A brief overview is given for two software developments related to the ITS code system. These developments provide parallel processing and visualization capabilities and thus allow users to perform ITS calculations more efficiently. Timing results and a graphical example are presented to demonstrate these capabilities

  8. A development and an application of Mixset-X computer code for simulating the Purex solvent extraction system

    International Nuclear Information System (INIS)

    Shida, M.; Naito, M.; Suto, T.; Omori, E.; Nojiri, T.

    2001-01-01

    MIXSET is a FORTRAN code developed to simulate the Purex solvent extraction system using mixer-settler extractors. Japan Nuclear Cycle Development Institute (JNC) has been developing the MIXSET code since the years 1970 to analyze the behavior of nuclides in the solvent extraction processes in Tokai Reprocessing Plant (TRP). This paper describes the history of MIXSET code development, the features of the latest version, called MIXSET-X and the application of the code for safety evaluation work. (author)

  9. Development of a coded aperture fuel motion diagnostics system for the ACPR (UPGRADE)

    International Nuclear Information System (INIS)

    Kelly, J.G.; Stalker, K.T.

    1979-01-01

    As part of Sandia Laboratories' program to study simulated core disruptive accidents in reactor safety research, a fuel motion detection system based on coded aperture imaging is being developed for the Annular Core Pulsed Reactor (ACPR). Although fuel motion has been observed at the TREAT by the fast neutron hodoscope and with a Vidicon pinhole camera technique, the coded aperture system offers a potential for lower cost, higher spatial resolution, three dimensional imaging, and higher frame rates at lower fluences than either of the other techniques

  10. Development of environmental dose assessment system (EDAS) code of PC version

    Energy Technology Data Exchange (ETDEWEB)

    Taki, Mitsumasa; Kikuchi, Masamitsu; Kobayashi, Hideo; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-05-01

    A computer code (EDAS) was developed to assess the public dose for the safety assessment to get the license of nuclear reactor operation. This code system is used for the safety analysis of public around the nuclear reactor in normal operation and severe accident. This code was revised and composed for personal computer user according to the Nuclear Safety Guidelines reflected the ICRP1990 recommendation. These guidelines are revised by Nuclear Safety Commission on March, 2001, which are 'Weather analysis guideline for the safety assessment of nuclear power reactor', 'Public dose around the facility assessment guideline corresponding to the objective value for nuclear power light water reactor' and 'Public dose assessment guideline for safety review of nuclear power light water reactor'. This code has been already opened for public user by JAERI, and English version code and user manual are also prepared. This English version code is helpful for international cooperation concerning the nuclear safety assessment with JAERI. (author)

  11. System transient analysis code development for low pressure and low power

    International Nuclear Information System (INIS)

    Kim, Hee Cheol

    1998-02-01

    A real time reactor system analysis code, ARTIST, based on drift flux model has been developed to investigate the transient system behavior under low pressure, low flow and low power conditions with noncondensable gas present in the system. The governing equations of the ARTIST code consist of three mass continuity equations (steam, liquid and noncondensable), two energy equations (gas and mixture) and one momentum equation (mixture) constituted with the drift flux model. The capability of ARTIST in predicting two-phase flow void distribution in the system has been validated against experimental data. The results of the ARTIST axial void distribution at low pressure and low flow, are far better than the results of both the homogeneous model of TASS code and the two-fluid model of RELAP5/MOD3 code. Also, RELAP5/MOD3 calculation shows the large amplitude of void fraction oscillations at low pressure. These results imply that interfacial momentum transfer terms in the two-fluid model formulation should be carefully constituted, especially for the low pressure condition due to the big density differences between steam and water. Thermal-hydraulic state solution scheme is developed when noncondensable gas exists. Numerical consistency and convergence of obtaining equilibrium state is tested with the ideal problems for various situations including very low partial pressure conditions. Calculated thermal-hydraulic state for each test shows consistent and expected behaviour. A new multi-layer back propagation network algorithm for calculating the departure from nucleate boiling ratio (DNBR) is developed and adopted in ARTIST code in order to have real-time DNBR evaluation by eliminating the tandem procedure of the transient DNBR calculation. The algorithm trained by different patterns generated by latin hypercube sampling method on the performance space is tested for the randomly sampled untrained data and the transient DNBR data. The uncertainty of the algorithm is

  12. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

  13. Study on a new meteorological sampling scheme developed for the OSCAAR code system

    International Nuclear Information System (INIS)

    Liu Xinhe; Tomita, Kenichi; Homma, Toshimitsu

    2002-03-01

    One important step in Level-3 Probabilistic Safety Assessment is meteorological sequence sampling, on which the previous studies were mainly related to code systems using the straight-line plume model and more efforts are needed for those using the trajectory puff model such as the OSCAAR code system. This report describes the development of a new meteorological sampling scheme for the OSCAAR code system that explicitly considers population distribution. A group of principles set for the development of this new sampling scheme includes completeness, appropriate stratification, optimum allocation, practicability and so on. In this report, discussions are made about the procedures of the new sampling scheme and its application. The calculation results illustrate that although it is quite difficult to optimize stratification of meteorological sequences based on a few environmental parameters the new scheme do gather the most inverse conditions in a single subset of meteorological sequences. The size of this subset may be as small as a few dozens, so that the tail of a complementary cumulative distribution function is possible to remain relatively static in different trials of the probabilistic consequence assessment code. (author)

  14. Development of An Automatic Verification Program for Thermal-hydraulic System Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W. [Pusan National University, Busan (Korea, Republic of); Suh, J. S.; Cho, Y. S.; Jeong, J. J. [System Engineering and Technology Co., Daejeon (Korea, Republic of)

    2012-05-15

    As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)

  15. Development of An Automatic Verification Program for Thermal-hydraulic System Codes

    International Nuclear Information System (INIS)

    Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W.; Suh, J. S.; Cho, Y. S.; Jeong, J. J.

    2012-01-01

    As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)

  16. Development of the next generation code system as an engineering modeling language (1)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; Ishikawa, Makoto; Ohira, Masanori; Kato, Masayuki

    2002-11-01

    In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenamine to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the physical properties and engineering models are represented as a programming language or a diagrams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language (EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML. (author)

  17. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  18. Development of system analysis code for pyrochemical process using molten salt electrorefining

    International Nuclear Information System (INIS)

    Tozawa, K.; Matsumoto, T.; Kakehi, I.

    2000-04-01

    This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basic of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without cooling gas circulation

  19. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    Homma, T.; Togawa, O.

    1991-01-01

    This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

  20. Development of system of computer codes for severe accident analysis and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Jang, H S; Jeon, M H; Cho, N J. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1992-01-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts.

  1. Development of system of computer codes for severe accident analysis and its applications

    International Nuclear Information System (INIS)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others

    1992-01-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts

  2. The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

    International Nuclear Information System (INIS)

    Wang Jianyu; Zhang Shenru; Min Yuanyou

    1994-01-01

    To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

  3. Reactor Systems Technology Division code development and configuration/quality control procedures

    International Nuclear Information System (INIS)

    Johnson, E.C.

    1985-06-01

    Procedures are prescribed for executing a code development task and implementing the resulting coding in an official version of a computer code. The responsibilities of the project manager, development staff members, and the Code Configuration/Quality Control Group are defined. Examples of forms, logs, computer job control language, and suggested outlines for reports associated with software production and implementation are included in Appendix A. 1 raf., 2 figs

  4. Comparative evaluation of various optimization methods and the development of an optimization code system SCOOP

    International Nuclear Information System (INIS)

    Suzuki, Tadakazu

    1979-11-01

    Thirty two programs for linear and nonlinear optimization problems with or without constraints have been developed or incorporated, and their stability, convergence and efficiency have been examined. On the basis of these evaluations, the first version of the optimization code system SCOOP-I has been completed. The SCOOP-I is designed to be an efficient, reliable, useful and also flexible system for general applications. The system enables one to find global optimization point for a wide class of problems by selecting the most appropriate optimization method built in it. (author)

  5. Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

  6. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  7. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    International Nuclear Information System (INIS)

    Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

    2010-10-01

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  8. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  10. Analyses and computer code developments for accident-induced thermohydraulic transients in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Wulff, W.

    1977-01-01

    A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment

  11. Development of the System Dynamics Code using Homogeneous Equilibrium Model for S-CO{sub 2} Brayton cycle Transient Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Seong Jun; Lee, Won Woong; Oh, Bongseong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The features of the S-CO{sub 2} Brayton cycle come from a small compressing work by designing the compressor inlet close the critical point of CO{sub 2}. This means the system condition can be operating under two-phase or sub-critical phase during transient situations such as changes of cooling system performance, load variations, etc. Since there is no operating MW scale S-CO{sub 2} Brayton cycle system in the world yet, using an analytical code is the only way to predict the system behavior and develop operating strategies of the S-CO{sub 2} Brayton cycles. Therefore, the development of a credible system code is an important part for the practical S-CO{sub 2} system research. The current status of the developed system analysis code for S-CO{sub 2} Brayton cycle transient analyses in KAIST and verification results are presented in this paper. To avoid errors related with convergences of the code during the phase changing flow calculation in GAMMA+ code, the authors have developed a system analysis code using Homogeneous Equilibrium Model (HEM) for the S-CO{sub 2} Brayton cycle transient analysis. The backbone of the in-house code is the GAMMA+1.0 code, but treating the quality of fluid by tracking system enthalpy gradient every time step. Thus, the code adopts pressure and enthalpy as the independent scalar variables to track the system enthalpy for updating the quality of the system every time step. The heat conduction solving method, heat transfer correlation and frictional losses on the pipe are referred from the GAMMA+ code.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  13. COCOSYS: Status of development and validation of the German containment code system

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Arndt, S.; Klein-Hessling, W.; Schwarz, S.; Spengler, C.; Weber, G.

    2006-01-01

    For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of a severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants. COCOSYS is composed for three main modules, which are separate executable files. The communication is realized via PVM (parallel virtual machine). The thermal hydraulic main module (THY) contains several specific models relevant for the simulation of severe accidents. Beside the usual capabilities to calculate the gas distribution and thermal behavior inside the containment, there are special models for the simulation of Hydrogen deflagration, pressure suppression systems etc. Further detailed models exist for the simulation of safety systems, like catalytic recombiners (PAR's), safety relief valves (used in WWR-440/V-230 type plants), ice condenser model, pump and spray system models for the complete simulation of cooling systems. The aerosol and fission product part (AFP) describes the aerosol behavior of nonsoluble and as well as hygroscopic aerosols, iodine chemistry and fission transport. Further the decay process of nuclides is considered using ORIGIN like routines. The corium concrete interaction (CCI) main module is based on an improved version of WECHSL extended by the ChemApp module for the

  14. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  15. Development of a thermal–hydraulic system code, TAPINS, for 10 MW regional energy reactor

    International Nuclear Information System (INIS)

    Lee, Yeon-Gun; Kim, Jong-Won; Park, Goon-Cherl

    2012-01-01

    Highlights: ► A thermal–hydraulic system code named TAPINS is developed for simulations of an integral reactor. ► The TAPINS is based on the one-dimensional momentum integral model. ► A dynamic model for the steam–gas pressurizer with non-condensable gas present is proposed. ► A series of pressurizer insurge test and natural circulation test are simulated by the TAPINS. ► It is proved that the TAPINS can provide reliable prediction of an integral reactor system on natural circulation. - Abstract: Small modular reactors (SMRs) with integral system layout have been drawing a great deal of attention as alternative options to branch out the utilization of nuclear energy as well as to offer the inherent safety features. Serving to confirm the design basis and analyze the transient behavior of an integral reactor such as REX-10, a thermal–hydraulic system code named TAPINS (Thermal–hydraulic Analysis Program for INtegral reactor System) is developed in this study. The TAPINS supports the simple pre-processing to build up the frameworks of node diagram for the typical integral reactor configuration. The TAPINS basically consists of mathematical models for the reactor coolant system, the core, the once-through helical-coil steam generator, and the built-in steam–gas pressurizer. The hydrodynamic model of the TAPINS is formulated using the one-dimensional momentum integral model, which is based on the analytical integration of the momentum equation around the closed loop in the system. As a key contribution of the study, a dynamic model for the steam–gas pressurizer with non-condensable gas present is newly proposed and incorporated into the code. The TAPINS is validated by comparing against the experimental data from the pressurizer insurge tests conducted at MIT (Massachusetts Institute of Technology) and natural circulation tests in the RTF (REX-10 Test Facility) at RERI (Regional Energy Reactor Institute). From the comparison results, it is

  16. Development of Integrated Code for Risk Assessment (INCORIA) for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Seo, Hyung Min; Yoo, Ho Sik

    2010-01-01

    A physical protection system (PPS) integrates people, procedures and equipment for the protection of assets or facilities against theft, sabotage or other malevolent human attacks. Among critical facilities, nuclear facilities and nuclear weapon sites require the highest level of PPS. After the September 11, 2001 terrorist attacks, international communities, including the IAEA, have made substantial efforts to protect nuclear material and nuclear facilities. The international flow on nuclear security is using the concept or risk assessment. The concept of risk assessment is firstly devised by nuclear safety people. They considered nuclear safety including its possible risk, which is the frequency of failure and possible consequence. Nuclear security people also considers security risk, which is the frequency of threat action, vulnerability, and consequences. The concept means that we should protect more when the credible threat exists and the possible radiological consequence is high. Even if there are several risk assessment methods of nuclear security, the application needs the help of tools because of a lot of calculation. It's also hard to find tools for whole phase of risk assessment. Several codes exist for the part of risk assessment. SAVI are used for vulnerability of PPS. Vital area identification code is used for consequence analysis. We are developing Integrated Code for Risk Assessment (INCORIA) to apply risk assessment methods for nuclear facilities. INCORIA evaluates PP-KINAC measures and generation tools for threat scenario. PP-KINAC is risk assessment measures for physical protection system developed by Hosik Yoo and is easy to apply. A threat scenario tool is used to generate threat scenario, which is used as one of input value to PP-KINAC measures

  17. Development of intelligent code system to support conceptual design of nuclear reactor core

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki; Tsuchihashi, Keichiro

    1997-01-01

    An intelligent reactor design system IRDS has been developed to support conceptual design of new type reactor cores in the fields of neutronics, thermal-hydraulics and fuel behavior. The features of IRDS are summarized as follows: 1) a variety of computer codes to cover various design tasks relevant to 'static' and 'burnup' problems are implemented, 2) all the information necessary to the codes implemented is unified in a data base, 3) several data and knowledge bases are referred to in order to proceed design process efficiently for non-expert users, 4) advanced man-machine interface to communicate with the system through an interactive and graphical user interface is equipped and 5) a function to search automatically a design window, which is defined as a feasible parameter range to satisfy design requirement and criteria is employed to support the optimization or satisfication process. Applicability and productivity of the system are demonstrated by the design study of fuel pin for new type FBR cores. (author)

  18. Preliminary Development of the MARS/FREK Spatial Kinetics Coupled System Code for Square Fueled Fast Reactor Applications

    International Nuclear Information System (INIS)

    Bae, Moo Hoon; Joo, Han Gyu

    2009-01-01

    Incorporation of a three-dimensional (3-D) reactor kinetics model into a system thermal-hydraulic (T/H) code enhances the capability to perform realistic analyses of the core neutronic behavior and the plant system dynamics which are coupled each other. For this advantage, several coupled system T/H and spatial kinetics codes, such as RELAP/PARCS, RELAP5/ PANBOX, and MARS/MASTER have been developed. These codes, however, so far limited to LWR applications. The objective of this work is to develop such a coupled code for fast reactor applications. Particularly, applications to lead-bismuth eutectic (LBE) cooled fast reactor are of interest which employ open square lattices. A fast reactor kinetics code applicable to square fueled cores called FREK is coupled the LBE version of the MARS code. The MARS/MASTER coupled code is used as the reference for the integration. The coupled code MARS/FREK is examined for a conceptual reactor called P-DEMO which is being developed by NUTRECK. In order to check the validity of the coupled code, however, the OECD MSLB benchmark exercise III calculation is solved first

  19. Development of the fire PSA methodology and the fire analysis computer code system

    International Nuclear Information System (INIS)

    Katsunori, Ogura; Tomomichi, Ito; Tsuyoshi, Uchida; Yusuke, Kasagawa

    2009-01-01

    Fire PSA methodology has been developed and was applied to NPPs in Japan for power operation and LPSD states. CDFs of preliminary fire PSA for power operation were the higher than that of internal events. Fire propagation analysis code system (CFAST/FDS Network) was being developed and verified thru OECD-PRISME Project. Extension of the scope for LPSD state is planned to figure out the risk level. In order to figure out the fire risk level precisely, the enhancement of the methodology is planned. Verification and validation of phenomenological fire propagation analysis code (CFAST/FDS Network) in the context of Fire PSA. Enhancement of the methodology such as an application of 'Electric Circuit Analysis' in NUREG/CR-6850 and related tests in order to quantify the hot-short effect precisely. Development of seismic-induced fire PSA method being integration of existing seismic PSA and fire PSA methods is ongoing. Fire PSA will be applied to review the validity of fire prevention and mitigation measures

  20. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    Reocreux, M.; Gauvain, J.

    1992-01-01

    The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab

  1. Development of platform to compare different wall heat transfer packages for system analysis codes

    International Nuclear Information System (INIS)

    Kim, Min-Gil; Lee, Won Woong; Lee, Jeong Ik; Shin, Sung Gil

    2016-01-01

    System thermal hydraulic (STH) analysis code is used for analyzing and evaluating the safety of a designed nuclear system. The system thermal hydraulic analysis code typically solves mass, momentum and energy conservation equations for multiple phases with sets of selected empirical constitutive equations to close the problem. Several STH codes are utilized in academia, industry and regulators, such as MARS-KS, SPACE, RELAP5, COBRA-TF, TRACE, and so on. Each system thermal hydraulic code consists of different sets of governing equations and correlations. However, the packages and sets of correlations of each code are not compared quantitatively yet. Wall heat transfer mode transition maps of SPACE and MARS-KS have a little difference for the transition from wall nucleate heat transfer mode to wall film heat transfer mode. Both codes have the same heat transfer packages and correlations in most region except for wall film heat transfer mode. Most of heat transfer coefficients calculated for the range of selected variables of SPACE are the same with those of MARS-KS. For the intervals between 500K and 540K of wall temperature, MARS-KS selects the wall film heat transfer mode and Bromley correlation but SPACE select the wall nucleate heat transfer mode and Chen correlation. This is because the transition from nucleate boiling to film boiling of MARS-KS is earlier than SPACE. More detailed analysis of the heat transfer package and flow regime package will be followed in the near future

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

  4. Revised SRAC code system

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.

    1986-09-01

    Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)

  5. Research on the improvement of nuclear safety -Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Kim, Dong Ha; Park, Won Seok; Hwang, Mi Jeong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated. These results will give a physical insight in the development of a new dispersion model. A wind tunnel experiment with bell shaped hill model was made in order to develop a new dispersion model. And an improved dispersion model was developed based on the concentration distribution data obtained from the wind tunnel experiment. This model will be added as an option to the atmospheric dispersion code. A stand-alone atmospheric code using MS Visual Basic programming language which runs at the Windows environment of a PC was developed. A user can easily select a necessary data file and type input data by clicking menus, and can select calculation options such building wake, plume rise etc., if necessary. And a user can easily understand the meaning of concentration distribution on the map around the plant site as well as output files. Also the methodologies for the estimation of radiation exposure and for the calculation of risks was established. These methodologies will be used for the development of modules for the radiation exposure and risks respectively. These modules will be developed independently and finally will be combined to the atmospheric dispersion code in order to develop a level 3 PSA code. 30 tabs., 56 figs., refs. (Author).

  6. Research on the improvement of nuclear safety -Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Kim, Dong Ha; Park, Won Seok; Hwang, Mi Jeong

    1995-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated. These results will give a physical insight in the development of a new dispersion model. A wind tunnel experiment with bell shaped hill model was made in order to develop a new dispersion model. And an improved dispersion model was developed based on the concentration distribution data obtained from the wind tunnel experiment. This model will be added as an option to the atmospheric dispersion code. A stand-alone atmospheric code using MS Visual Basic programming language which runs at the Windows environment of a PC was developed. A user can easily select a necessary data file and type input data by clicking menus, and can select calculation options such building wake, plume rise etc., if necessary. And a user can easily understand the meaning of concentration distribution on the map around the plant site as well as output files. Also the methodologies for the estimation of radiation exposure and for the calculation of risks was established. These methodologies will be used for the development of modules for the radiation exposure and risks respectively. These modules will be developed independently and finally will be combined to the atmospheric dispersion code in order to develop a level 3 PSA code. 30 tabs., 56 figs., refs. (Author)

  7. Development of OCDMA system based on Flexible Cross Correlation (FCC) code with OFDM modulation

    Science.gov (United States)

    Aldhaibani, A. O.; Aljunid, S. A.; Anuar, M. S.; Arief, A. R.; Rashidi, C. B. M.

    2015-03-01

    The performance of the OCDMA systems is governed by numerous quantitative parameters such as the data rate, simultaneous number of users, the powers of transmitter and receiver, and the type of codes. This paper analyzes the performance of the OCDMA system using OFDM technique to enhance the channel data rate, to save power and increase the number of user of OSCDMA systems compared with previous hybrid subcarrier multiplexing/optical spectrum code division multiplexing (SCM/OSCDM) system. The average received signal to noise ratio (SNR) with the nonlinearity of subcarriers is derived. The theoretical results have been evaluated based on BER and number of users as well as amount of power saved. The proposed system gave better performance and save around -6 dBm of the power as well as increase the number of users twice compare to SCM/OCDMA system. In addition it is robust against interference and much more spectrally efficient than SCM/OCDMA system. The system was designed based on Flexible Cross Correlation (FCC) code which is easier construction, less complexity of encoder/decoder design and flexible in-phase cross-correlation for uncomplicated to implement using Fiber Bragg Gratings (FBGs) for the OCDMA systems for any number of users and weights. The OCDMA-FCC_OFDM improves the number of users (cardinality) 108% compare to SCM/ODCMA-FCC system.

  8. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-04-01

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

  9. ETR/ITER systems code

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  10. Development of the DTNTES code

    International Nuclear Information System (INIS)

    Ortega Prieto, P.; Morales Dorado, M.D.; Alonso Santos, A.

    1987-01-01

    The DTNTES code has been developed in the Department of Nuclear Technology of the Polytechnical University in Madrid as a part of the Research Program on Quantitative Risk Analysis. DTNTES code calculates several time-dependent probabilistic characteristics of basic events, minimal cut sets and the top event of a fault tree. The code assumes that basic events are statistically independent, and they have failure and repair distributions. It computes the minimal cut upper bound approach for the top event unavailability, and the time-dependent unreliability of the top event by means of different methods, selected by the user. These methods are: expected number of system failures, failure rate, Barlow-Proschan bound, steady-state upper bound, and T* method. (author)

  11. Development of a new flux map processing code for moveable detector system in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Li, W.; Lu, H.; Li, J.; Dang, Z.; Zhang, X. [China Nuclear Power Technology Research Institute, 47 F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen 518026 (China); Wu, Y.; Fan, X. [Information Technology Center, China Guangdong Nuclear Power Group, Shenzhen 518000 (China)

    2013-07-01

    This paper presents an introduction to the development of the flux map processing code MAPLE developed by China Nuclear Power Technology Research Institute (CNPPJ), China Guangdong Nuclear Power Group (CGN). The method to get the three-dimensional 'measured' power distribution according to measurement signal has also been described. Three methods, namely, Weight Coefficient Method (WCM), Polynomial Expand Method (PEM) and Thin Plane Spline (TPS) method, have been applied to fit the deviation between measured and predicted results for two-dimensional radial plane. The measured flux map data of the LINGAO nuclear power plant (NPP) is processed using MAPLE as a test case to compare the effectiveness of the three methods, combined with a 3D neutronics code COCO. Assembly power distribution results show that MAPLE results are reasonable and satisfied. More verification and validation of the MAPLE code will be carried out in future. (authors)

  12. Development of a software for the ASME code qualification of class-I nuclear piping systems

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Umashankar, C.; Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

    1999-11-01

    In nuclear industry, the designer often comes across the requirements of Class-1 piping systems which need to be qualified for various normal and abnormal loading conditions. In order to have quick design changes and the design reviews at various stages of design, it is quite helpful if a dedicated software is available for the qualification of Class-1 piping systems. BARC has already purchased a piping analysis software CAESAR-II and has used it for the life extension of heavy water plant, Kota. CAESAR-II facilitates the qualification of Class-2 and Class-3 piping systems among others. However, the present version of CAESAR-II does not have the capability to perform stress checks for the ASME Class-1 nuclear piping systems. With this requirement in mind and the prohibitive costs of commercially available software for the Class-1 piping analyses, it was decided to develop a separate software for this class of piping in such a way that the input and output details of the piping from the CAESAR-II software can be made use of. This report principally contains the details regarding development of a software for codal qualification of Class-1 nuclear piping as per ASME code section-III, NB-3600. The entire work was carried out in three phases. The first phase consisted of development of the routines for reading the output files obtained from the CAESAR-II software, and converting them into required format for further processing. In this phase, the nodewise informations available from the CAESAR-II output file were converted into element-wise informations. The second phase was to develop a general subroutine for reading the various input parameters such as diameter, wall thickness, corrosion allowance, bend radius and also to recognize the bend elements based on the bend radius, directly from the input file of CAESAR-II software. The third phase was regarding the incorporation of the required steps for performing the ASME codal checks as per NB-3600 for Class-1 piping

  13. Recent activities in accelerator code development

    International Nuclear Information System (INIS)

    Copper, R.K.; Ryne, R.D.

    1992-01-01

    In this paper we will review recent activities in the area of code development as it affects the accelerator community. We will first discuss the changing computing environment. We will review how the computing environment has changed in the last 10 years, with emphasis on computing power, operating systems, computer languages, graphics standards, and massively parallel processing. Then we will discuss recent code development activities in the areas of electromagnetics codes and beam dynamics codes

  14. Development of code system for management of reactor decommissioning (COSMARD), 1

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi; Ogihara, Hirohito

    1994-02-01

    The Code System for Management of Reactor Decommissioning (COSMARD) was developed for use in the effective planning and management of reactor decommissioning. The decommissioning management data evaluation facility (DMAF) which is the main part of COSMARD has functions to evaluate various project management data such as manpower needs, radiation exposure of workers, amount of waste arisings necessary for each activity in a project using input data and calculation models consisting of simple arithmetic formulas and unit factors in the database. Using a set of command descriptors developed in COSMARD, work conditions and procedures for decommissioning a nuclear facility are describes as input data. The management data are evaluated by adopting the calculation models, which are placed in the activities at the lowest level of the work breakdown structure (WBS). The management data evaluated by the models are summed up in the ascending direction of WBS to obtain necessary data for the activities at any levels of WBS. In addition, scheduling calculations are conducted to obtain scheduling bar chart and histograms of the management data, on the basis of the work precedence conditions attached at certain activities. This report describes the outline of DMAF and user's manual of the sets of command descriptors. (author)

  15. [Development of the 2014 G-DRG system. Departure from coding of secondary diagnoses?].

    Science.gov (United States)

    Volkmer, B G; Kahlmeyer, A; Petervari, M; Pechoel, M

    2014-01-01

    The objective of the German DRG (diagnosis-related groups) system is to adequately reimburse hospital costs using flat rate payments. The goal is to thereby achieve the most adequate representation of hospital costs in flat rate payments. The DRG for 2014 is based on the actual number of cases treated and the costs determined from 2012. For 2014, the current changes of the DRG system for the specialty urology concerning the coding and recording of secondary diagnoses are presented and discussed.

  16. Development of a system code for transient analysis in a HTGR

    International Nuclear Information System (INIS)

    Lee, Tae Beom

    2004-02-01

    A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point

  17. Recent developments and improvements in the code system for the neutronic design of fast breeder reactors at CEA

    International Nuclear Information System (INIS)

    Giacometti; Estiot, J.C.; Palmiotti, G.; Grondein, C.; Le Cardinal, G.; Ravier, M.

    1982-09-01

    The new modular system has been developed using software tools ALOS, which give effectiveness and flexibility. Second step is the application of new algorithms to improve the resolution of diffusion equation. The state of art for the realisation of the new code system is briefly described

  18. Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

  19. Development of computer code in PNC, 3

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ohira, Hiroaki

    1990-01-01

    Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

  20. Development of a computational system based in the code GEANT4 for dosimetric evaluation in radiotherapy

    International Nuclear Information System (INIS)

    Oliveira, Alex Cristovao Holanda de

    2016-01-01

    The incidence of cancer has grown in Brazil, as well as around the world, following the change in the age profile of the population. One of the most important techniques and commonly used in cancer treatment is radiotherapy. Around 60% of new cases of cancer use radiation in at least one phase of treatment. The most used equipment for radiotherapy is a linear accelerator (Linac) which produces electron or X-ray beams in energy range from 5 to 30 MeV. The most appropriate way to irradiate a patient is determined during treatment planning. Currently, treatment planning system (TPS) is the main and the most important tool in the process of planning for radiotherapy. The main objective of this work is to develop a computational system based on the MC code Geant4 for dose evaluations in photon beam radiotherapy. In addition to treatment planning, these dose evaluations can be performed for research and quality control of equipment and TPSs. The computer system, called Quimera, consists of a graphical user interface (qGUI) and three MC applications (qLinacs, qMATphantoms and qNCTphantoms). The qGUI has the function of interface for the MC applications, by creating or editing the input files, running simulations and analyzing the results. The qLinacs is used for modeling and generation of Linac beams (phase space). The qMATphantoms and qNCTphantoms are used for dose calculations in virtual models of physical phantoms and computed tomography (CT) images, respectively. From manufacturer's data, models of a Varian Linac photon beam and a Varian multileaf collimator (MLC) were simulated in the qLinacs. The Linac and MLC modelling were validated using experimental data. qMATphamtoms and qNCTphantoms were validated using IAEA phase spaces. In this first version, the Quimera can be used for research, radiotherapy planning of simple treatments and quality control in photon beam radiotherapy. The MC applications work independent of the qGUI and the qGUI can be used for

  1. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

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

  2. Development of Attendance Database System Using Bar-coded Student Card

    Directory of Open Access Journals (Sweden)

    Abdul Fadlil

    2008-04-01

    Full Text Available The calculation of the level of attendance is very important, because one indicator of a person's credibility can be seen from the level of attendance. For example, at a university, data about the level of attendance of a student in a lecture is very important as one of components in the assessment. The manual presence system is considered less effective. This research presents the draft of presence system using bar codes (barcodes as input data representing the attendance. The presence system is supported by three main components, those are a bar code found on the student card (KTM, a CCD barcode scanner series and a CD-108E computer. Management of attendance list using this system allows for optimization of functions of KTM. The presence system has been tested with several KTM through a variety of distances and positions of the barcode scanner barcode. The test results is obtained at ideal position for reading a barcode when a barcode scanner is at 2 cm from the object with 90 degree. At this position the level of accuracy reach 100%.

  3. Development status of TUF code

    International Nuclear Information System (INIS)

    Liu, W.S.; Tahir, A.; Zaltsgendler

    1996-01-01

    An overview of the important development of the TUF code in 1995 is presented. The development in the following areas is presented: control of round-off error propagation, gas resolution and release models, and condensation induced water hammer. This development is mainly generated from station requests for operational support and code improvement. (author)

  4. SCALE Code System

    Energy Technology Data Exchange (ETDEWEB)

    Jessee, Matthew Anderson [ORNL

    2016-04-01

    The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE

  5. Overview of Development and Deployment of Codes, Standards and Regulations Affecting Energy Storage System Safety in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Conover, David R.

    2014-08-22

    This report acquaints stakeholders and interested parties involved in the development and/or deployment of energy storage systems (ESS) with the subject of safety-related codes, standards and regulations (CSRs). It is hoped that users of this document gain a more in depth and uniform understanding of safety-related CSR development and deployment that can foster improved communications among all ESS stakeholders and the collaboration needed to realize more timely acceptance and approval of safe ESS technology through appropriate CSR.

  6. Preliminary investigation study of code of developed country for developing Korean fuel cycle code

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2012-01-01

    In order to develop Korean fuel cycle code, the analyses has been performed with the fuel cycle codes which are used in advanced country. Also, recommendations were proposed for future development. The fuel cycle codes are AS FLOOWS: VISTA which has been developed by IAEA, DANESS code which developed by ANL and LISTO, and VISION developed by INL for the Advanced Fuel Cycle Initiative (AFCI) system analysis. The recommended items were proposed for software, program scheme, material flow model, isotope decay model, environmental impact analysis model, and economics analysis model. The described things will be used for development of Korean nuclear fuel cycle code in future

  7. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  8. Embedded Systems Hardware Integration and Code Development for Maraia Capsule and E-MIST

    Science.gov (United States)

    Carretero, Emmanuel S.

    2015-01-01

    The cost of sending large spacecraft to orbit makes them undesirable for carrying out smaller scientific missions. Small spacecraft are more economical and can be tailored for missions where specific tasks need to be carried out, the Maraia capsule is such a spacecraft. Maraia will allow for samples of experiments conducted on the International Space Station to be returned to earth. The use of balloons to conduct experiments at the edge of space is a practical approach to reducing the large expense of using rockets. E-MIST is a payload designed to fly on a high altitude balloon. It can maintain science experiments in a controlled manner at the edge of space. The work covered here entails the integration of hardware onto each of the mentioned systems and the code associated with such work. In particular, the resistance temperature detector, pressure transducers, cameras, and thrusters for Maraia are discussed. The integration of the resistance temperature detectors and motor controllers to E-MIST is described. Several issues associated with sensor accuracy, code lock-up, and in-flight reset issues are mentioned. The solutions and proposed solutions to these issues are explained.

  9. Status of the development of a fully integrated code system for the simulation of high temperature reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Kasselmann, Stefan, E-mail: s.kasselmann@fz-juelich.de [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Druska, Claudia [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Herber, Stefan [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Jühe, Stephan [Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Keller, Florian; Lambertz, Daniela; Li, Jingjing; Scholthaus, Sarah; Shi, Dunfu [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Xhonneux, Andre; Allelein, Hans-Josef [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany)

    2014-05-01

    The HTR code package (HCP) is a new code system, which couples a variety of stand-alone codes for the simulation of different aspects of HTR. HCP will allow the steady-state and transient operating conditions of a 3D reactor core to be simulated including new features such as spatially resolved fission product release calculations or production and transport of graphite dust. For this code the latest programming techniques and standards are applied. As a first step an object-oriented data model was developed which features a high level of readability because it is based on problem-specific data types like Nuclide, Reaction, ReactionHandler, CrossSectionSet, etc. Those classes help to encapsulate and therefore hide specific implementations, which are not relevant with respect to physics. HCP will make use of one consistent data library for which an automatic generation tool was developed. The new data library consists of decay information, cross sections, fission yields, scattering matrices etc. for all available nuclides (e.g. ENDF/B-VII.1). The data can be stored in different formats such as binary, ASCII or XML. The new burn up code TNT (Topological Nuclide Transmutation) applies graph theory to represent nuclide chains and to minimize the calculation effort when solving the burn up equations. New features are the use of energy-dependent fission yields or the calculation of thermal power for decay, fission and capture reactions. With STACY (source term analysis code system) the fission product release for steady state as well as accident scenarios can be simulated for each fuel batch. For a full-core release calculation several thousand fuel elements are tracked while passing through the core. This models the stochastic behavior of a pebble bed in a realistic manner. In this paper we report on the current status of the HCP and present first results, which prove the applicability of the selected approach.

  10. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  11. Development and assessment of Multi-dimensional flow models in the thermal-hydraulic system analysis code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M

    2005-04-15

    A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST.

  12. Development and assessment of Multi-dimensional flow models in the thermal-hydraulic system analysis code MARS

    International Nuclear Information System (INIS)

    Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M.

    2005-04-01

    A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST

  13. Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

    International Nuclear Information System (INIS)

    Hong, B.G.; Lee, Dong Won; Kim, Yong Hi

    2007-01-01

    To develop the concepts of fusion power plants and identify the design parameters, we have been developing the tokamak reactor system code. The system code can take into account a wide range of plasma physics and technology effects simultaneously and it can be used to find design parameters which optimize the given figure of merits. The outcome of the system studies using the system code is to identify which areas of plasma physics and technologies and to what extent should be developed for realization of a given fusion power plant concepts. As an application of the tokamak reactor system code, we investigate the performance of DEMO for early realization with a limited extension from the plasma physics and technology used in the design of the ITER. Main requirements for DEMO are selected as: 1) to demonstrate tritium self-sufficiency, 2) to generate net electricity, and 3) for steady-state operation. The size of plasma is assumed to be same as that of ITER and the plasma parameters which characterize the performance, i.e. normalized β value, β N , confinement improvement factor for the H-mode, H and the ratio of the Greenwald density limit n/n G are assumed to be improved beyond those of ITER: β N >2.0, H>1.0 and n/n G >1.0. Tritium self-sufficiency is provided by the He Cooled Molten Lithium (HCML) blanket with the total thickness of 2.5 m including the shield. With n/n G >1.2, net electric power bigger than 500 MW is possible with β N >4.0 andH>1.2. To access operation space for higher electric power, main restrictions are given by the divertor heat load and the steady-state operation requirements. Developments in both plasma physics and technology are required to handle high heat load and to increase the current drive efficiency. (orig.)

  14. UUGM Code Development

    Science.gov (United States)

    1993-07-26

    atmospheres of at this time. certain solar system planets. The JPL work was directed at very specific applications; however, the Studi"ss-1 7 WorkRe ported in...pressure atmospheres of certain 1-60 Hz. The spark plug power output was varied inversely solar system planets. The JPL work was directed at very with...general case of explosive dust diaper - deposited as a thin layer over the surface area of the sion. We will consider particle density variation from 4

  15. LiveCode mobile development

    CERN Document Server

    Lavieri, Edward D

    2013-01-01

    A practical guide written in a tutorial-style, ""LiveCode Mobile Development Hotshot"" walks you step-by-step through 10 individual projects. Every project is divided into sub tasks to make learning more organized and easy to follow along with explanations, diagrams, screenshots, and downloadable material.This book is great for anyone who wants to develop mobile applications using LiveCode. You should be familiar with LiveCode and have access to a smartphone. You are not expected to know how to create graphics or audio clips.

  16. Development of a 14-digit Hydrologic Unit Code Numbering System for South Carolina

    Science.gov (United States)

    Bower, David E.; Lowry, Claude; Lowery, Mark A.; Hurley, Noel M.

    1999-01-01

    A Hydrologic Unit Map showing the cataloging units, watersheds, and subwatersheds of South Carolina has been developed by the U.S. Geological Survey in cooperation with the South Carolina Department of Health and Environmental Control, funded through a U.S. Environmental Protection Agency 319 Grant, and the U.S. Department of Agriculture, Natural Resources Conservation Service. These delineations represent 8-, 11-, and 14-digit Hydrologic Unit Codes, respectively. This map presents information on drainage, hydrography, and hydrologic boundaries of the water-resources regions, subregions, accounting units, cataloging units, watersheds, and subwatersheds. The source maps for the basin delineations are 1:24,000-scale 7.5-minute series topographic maps and the base maps are from 1:100,000-scale Digital Line Graphs; however, the data are published at a scale of 1:500,000. In addition, an electronic version of the data is provided on a compact disc.Of the 1,022 subwatersheds delineated for this project, 1,004 range in size from 3,000 to 40,000 acres (4.69 to 62.5 square miles). Seventeen subwatersheds are smaller than 3,000 acres and one subwatershed, located on St. Helena Island, is larger than 40,000 acres.This map and its associated codes provide a standardized base for use by water-resource managers and planners in locating, storing, retrieving, and exchanging hydrologic data. In addition, the map can be used for cataloging water-data acquisition activities, geographically organizing hydrologic data, and planning and describing water-use and related land-use activities.

  17. Elements of algebraic coding systems

    CERN Document Server

    Cardoso da Rocha, Jr, Valdemar

    2014-01-01

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

  18. Coarse mesh code development

    Energy Technology Data Exchange (ETDEWEB)

    Lieberoth, J.

    1975-06-15

    The numerical solution of the neutron diffusion equation plays a very important role in the analysis of nuclear reactors. A wide variety of numerical procedures has been proposed, at which most of the frequently used numerical methods are fundamentally based on the finite- difference approximation where the partial derivatives are approximated by the finite difference. For complex geometries, typical of the practical reactor problems, the computational accuracy of the finite-difference method is seriously affected by the size of the mesh width relative to the neutron diffusion length and by the heterogeneity of the medium. Thus, a very large number of mesh points are generally required to obtain a reasonably accurate approximate solution of the multi-dimensional diffusion equation. Since the computation time is approximately proportional to the number of mesh points, a detailed multidimensional analysis, based on the conventional finite-difference method, is still expensive even with modern large-scale computers. Accordingly, there is a strong incentive to develop alternatives that can reduce the number of mesh-points and still retain accuracy. One of the promising alternatives is the finite element method, which consists of the expansion of the neutron flux by piecewise polynomials. One of the advantages of this procedure is its flexibility in selecting the locations of the mesh points and the degree of the expansion polynomial. The small number of mesh points of the coarse grid enables to store the results of several of the least outer iterations and to calculate well extrapolated values of them by comfortable formalisms. This holds especially if only one energy distribution of fission neutrons is assumed for all fission processes in the reactor, because the whole information of an outer iteration is contained in a field of fission rates which has the size of all mesh points of the coarse grid.

  19. Development of the Chronic Pain Coding System (CPCS) for Characterizing Patient-Clinician Discussions About Chronic Pain and Opioids.

    Science.gov (United States)

    Henry, Stephen G; Chen, Meng; Matthias, Marianne S; Bell, Robert A; Kravitz, Richard L

    2016-10-01

    To describe the development and initial application of the Chronic Pain Coding System. Secondary analysis of data from a randomized clinical trial. Six primary care clinics in northern California. Forty-five primary care visits involving 33 clinicians and 45 patients on opioids for chronic noncancer pain. The authors developed a structured coding system to accurately and objectively characterize discussions about pain and opioids. Two coders applied the final system to visit transcripts. Intercoder agreement for major coding categories was moderate to substantial (kappa = 0.5-0.7). Mixed effects regression was used to test six hypotheses to assess preliminary construct validity. Greater baseline pain interference was associated with longer pain discussions (P = 0.007) and more patient requests for clinician action (P = 0.02) but not more frequent negative patient evaluations of pain (P = 0.15). Greater clinician-reported visit difficulty was associated with more frequent disagreements with clinician recommendations (P = 0.003) and longer discussions of opioid risks (P = 0.049) but not more frequent requests for clinician action (P = 0.11). Rates of agreement versus disagreement with patient requests and clinician recommendations were similar for opioid-related and non-opioid-related utterances. This coding system appears to be a reliable and valid tool for characterizing patient-clinician communication about opioids and chronic pain during clinic visits. Objective data on how patients and clinicians discuss chronic pain and opioids are necessary to identify communication patterns and strategies for improving the quality and productivity of discussions about chronic pain that may lead to more effective pain management and reduce inappropriate opioid prescribing. © 2016 American Academy of Pain Medicine. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  20. System Based Code: Principal Concept

    International Nuclear Information System (INIS)

    Yasuhide Asada; Masanori Tashimo; Masahiro Ueta

    2002-01-01

    This paper introduces a concept of the 'System Based Code' which has initially been proposed by the authors intending to give nuclear industry a leap of progress in the system reliability, performance improvement, and cost reduction. The concept of the System Based Code intends to give a theoretical procedure to optimize the reliability of the system by administrating every related engineering requirement throughout the life of the system from design to decommissioning. (authors)

  1. Development of the next generation code system as an engineering modeling language (6). Development of a cross section adjustment and nuclear design accuracy evaluation solver

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2008-01-01

    A new cross section adjustment and nuclear design accuracy evaluation solver was developed as a set of modules for MARBLE (multi-purpose advanced reactor physics analysis system based on language of engineering). In order to enhance the system extendibility and flexibility, the object-oriented design and analysis technique was adopted to the development. In the new system, it is easy to add a new design accuracy evaluation method because a new numerical calculation module is independent from other modules. Further, several new functions such as searching and editing calculation data are provided in the new solver. These functions can be easily customised by users because they are designed to work cooperatively with Python scripting language, which is used as a user interface of the MARBLE system. In order to validate the new solver, a test calculation was performed for a realistic calculation case of creating a new unified cross section library. In the test calculation, results calculated by the new solver agreed well with those by the conventional code system. In addition, it is possible to reuse existing input data files prepared for the conventional code system because the new solver utilities support the conventional formats. Because the new solver implements all main functions of the conventional code system, MARBLE can be used as a new calculation code system for cross section adjustment and nuclear design accuracy evaluation

  2. Real-space grids and the Octopus code as tools for the development of new simulation approaches for electronic systems

    Science.gov (United States)

    Andrade, Xavier; Strubbe, David; De Giovannini, Umberto; Larsen, Ask Hjorth; Oliveira, Micael J. T.; Alberdi-Rodriguez, Joseba; Varas, Alejandro; Theophilou, Iris; Helbig, Nicole; Verstraete, Matthieu J.; Stella, Lorenzo; Nogueira, Fernando; Aspuru-Guzik, Alán; Castro, Alberto; Marques, Miguel A. L.; Rubio, Angel

    Real-space grids are a powerful alternative for the simulation of electronic systems. One of the main advantages of the approach is the flexibility and simplicity of working directly in real space where the different fields are discretized on a grid, combined with competitive numerical performance and great potential for parallelization. These properties constitute a great advantage at the time of implementing and testing new physical models. Based on our experience with the Octopus code, in this article we discuss how the real-space approach has allowed for the recent development of new ideas for the simulation of electronic systems. Among these applications are approaches to calculate response properties, modeling of photoemission, optimal control of quantum systems, simulation of plasmonic systems, and the exact solution of the Schr\\"odinger equation for low-dimensionality systems.

  3. Development of Tokamak Reactor System Code and Performance for Early Realization of DEMO

    International Nuclear Information System (INIS)

    Hong, B. G.; Lee, D. W.; Kim, Y.

    2006-01-01

    To develop the concepts of DEMO and identify the design parameters, dependence on performance objectives, design features and physical and technical constraints have to be considered. System analyses are necessary to find device variables which optimize figures of merit such as major radius, ignition margin, divertor heat load, neutron wall load, etc. Demonstration fusion power plant, DEMO is regarded as the last step before the development of a commercial fusion reactor in Korea National Basic Plan for the Development of Fusion Energy. The DEMO should demonstrate a net electric power generation, a tritium self sufficiency, and the safety aspect of a power plant. Performance of DEMO for early realization has been investigated with a limited extension from the plasma physics and technology in the 2nd phase of the ITER operation (EPP phase)

  4. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

    International Nuclear Information System (INIS)

    Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

    2012-01-01

    By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

  5. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

    International Nuclear Information System (INIS)

    Yang, W.S.; Lee, C.H.

    2008-01-01

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

  6. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

    2008-05-16

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

  7. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

    International Nuclear Information System (INIS)

    Campolina, Daniel de Almeida Magalhaes

    2009-01-01

    In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

  8. Development of TIME2 code

    International Nuclear Information System (INIS)

    1986-02-01

    The paper reviews the progress on the development of a computer model TIME2, for modelling the long term evolution of shallow burial site environments for low- and intermediate-level radioactive waste disposal. The subject is discussed under the five topic headings: 1) background studies, including geomorphology, climate, human-induced effects, and seismicity, 2) development of the TIME2 code, 3) verification and testing, 4) documentation, and, 5) role of TIME2 in radiological risk assessment. (U.K.)

  9. Development of HTGR plant dynamics simulation code

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Tazawa, Yujiro; Mitake, Susumu; Suzuki, Katsuo.

    1987-01-01

    Plant dynamics simulation analysis plays an important role in the design work of nuclear power plant especially in the plant safety analysis, control system analysis, and transient condition analysis. The authors have developed the plant dynamics simulation code named VESPER, which is applicable to the design work of High Temperature Engineering Test Reactor, and have been improving the code corresponding to the design changes made in the subsequent design works. This paper describes the outline of VESPER code and shows its sample calculation results selected from the recent design work. (author)

  10. Computer code development plant for SMART design

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  11. Computer code development plant for SMART design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  12. Multiple application coded switch development report

    International Nuclear Information System (INIS)

    Bernal, E.L.; Kestly, J.D.

    1979-03-01

    The development of the Multiple Application Coded Switch (MACS) and its related controller are documented; the functional and electrical characteristics are described; the interface requirements defined, and a troubleshooting guide provided. The system was designed for the Safe Secure Trailer System used for secure transportation of nuclear material

  13. SPECTRAL AMPLITUDE CODING OCDMA SYSTEMS USING ENHANCED DOUBLE WEIGHT CODE

    Directory of Open Access Journals (Sweden)

    F.N. HASOON

    2006-12-01

    Full Text Available A new code structure for spectral amplitude coding optical code division multiple access systems based on double weight (DW code families is proposed. The DW has a fixed weight of two. Enhanced double-weight (EDW code is another variation of a DW code family that can has a variable weight greater than one. The EDW code possesses ideal cross-correlation properties and exists for every natural number n. A much better performance can be provided by using the EDW code compared to the existing code such as Hadamard and Modified Frequency-Hopping (MFH codes. It has been observed that theoretical analysis and simulation for EDW is much better performance compared to Hadamard and Modified Frequency-Hopping (MFH codes.

  14. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are building a new framework of common supporting utilities and software tools to facilitate further development. research and development on basic computational techniques in classical mechanics and electrodynamics, and evaluation and comparison of existing beam optics codes, and support for their continuing development

  15. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are: building a new framework of common supporting utilities and software tools to facilitate further development; research and development on basic computational techniques in classical mechanics and electrodynamics; and evaluation and comparison of existing beam optics codes, and support for their continuing development. 17 refs

  16. Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms

    International Nuclear Information System (INIS)

    Datta, Debashis

    2010-02-01

    A nuclear piping system is composed of several straight pipes and elbows joined by welding. These weld sections are usually the most susceptible failure parts susceptible to various degradation mechanisms. Whereas a specific location of a reactor piping system might fail by a combination of different aging mechanisms, e.g. fatigue and/or stress corrosion cracking, the majority of the piping probabilistic fracture mechanics (PFM) codes can only consider a single aging mechanism at a time. So, a probabilistic fracture mechanics computer code capable of considering multiple aging mechanisms was developed for an accurate failure analysis of each specific component of a nuclear piping section. The newly proposed crack morphology based probabilistic leak flow rate module is introduced in this code to separately treat fatigue and SCC type cracks. Improved models e.g. stressors models, elbow failure model, SIFs model, local seismic occurrence probability model, performance based crack detection models, etc., are also included in this code. Recent probabilistic fatigue (S-N) and SCC crack initiation (S-T) and subsequent crack growth rate models are coded. An integrated probabilistic risk assessment and probabilistic fracture mechanics methodology is proposed. A complete flow chart regarding the combined aging mechanism model is presented. The combined aging mechanism based module can significantly reduce simulation efforts and time. Two NUREG benchmark problems, e.g. reactor pressure vessel outlet nozzle section and a surge line elbow located just below the pressurizer are reinvestigated by this code. The results showed that, contribution of pre-existing cracks in addition to initiating cracks, can significantly increase the overall failure probability. Inconel weld location of reactor pressure vessel outlet nozzle section showed the weakest point in terms of relative through-wall leak failure probability in the order of about 10 -2 at the 40-year plant life. Considering

  17. Algorithmic developments and qualification of the ERANOS nuclear code and data system for the characterization of fast neutron reactors

    International Nuclear Information System (INIS)

    Rimpault, G.

    2003-09-01

    In this report, the author discusses the algorithmic and methodological developments in the field of nuclear reactor physics, and more particularly the developments of the ERALIB1/ERANOS nuclear code and data system for the calculation of core critical mass and power of sodium-cooled fast neutron reactors (Phenix and Super Phenix), and of the CAPRA 4/94 core. After a brief recall of nuclear data and methods used to determine critical masses and powers, the author discusses the interpretation of start-up experiments performed on Super-Phenix. The methodology used to characterize the uncertainties of these parameters is then applied to the calculation of the Super-Phenix critical mass and power distribution. He presents the approach chosen to define the validity domain of the ERANOS form

  18. Development of next generation code system as an engineering modeling language (5). Investigation on restructuring method of conventional codes into two-layer system

    International Nuclear Information System (INIS)

    Yokoyama, Kenji

    2006-10-01

    A proposed method for gradually restructuring to the two-level system of next generation analysis system by reusing the conventional analysis system, called 'incremental method', was applied and evaluated. The following functions were selected for the evaluation: Neutron diffusion calculation for the three-dimensional XYZ system based on finite differential method, and input utilities of the cross-section data file used in the conventional system. In order to evaluate the effect of the restructuring, 'Module Coupling Index (MCI)' and 'McCabe's Cyclomatic Complexity (MCC)' were used for quantifying the quality of the modular design and the complexity of the program sequence of each module. Although MCIs of each module before restructuring were mainly 6 - 7 degrees, it was possible to reduce them to under 4 degrees in most module by restructuring with the incremental method. And, it is found that the modules under 4 degrees of MCI can be easily combined with different programming languages, which are necessary for building the two-layer system. In the meantime, MCCs in most module before restructuring were over 20 and some were over 50. The incremental method could reduce them to under 10 when C++ was used, and reduce them to under 20 when FORTRAN was used. It is correspondent to reduction of the error frequency occurred in its modification from 20 - 40% to 5 - 10%. The total number of MCC could be reduced to 1/3 when C++ was used, and to 1/2 when FORTRAN was used. By using the restructured functions in the present study and some previously developed functions, a reactor analysis tool was systematized and applied to criticality analysis of the Experimental Fast Reactor 'JOYO' MR-I. In addition, the following two functionality expansion tests were performed: To add cross section direct perturbation functionality, and to add control rod criticality position search functionality. In the tests, both the functionality expansions were carried out satisfying the condition

  19. 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

  20. The development of code benchmarks

    International Nuclear Information System (INIS)

    Glass, R.E.

    1986-01-01

    Sandia National Laboratories has undertaken a code benchmarking effort to define a series of cask-like problems having both numerical solutions and experimental data. The development of the benchmarks includes: (1) model problem definition, (2) code intercomparison, and (3) experimental verification. The first two steps are complete and a series of experiments are planned. The experiments will examine the elastic/plastic behavior of cylinders for both the end and side impacts resulting from a nine meter drop. The cylinders will be made from stainless steel and aluminum to give a range of plastic deformations. This paper presents the results of analyses simulating the model's behavior using materials properties for stainless steel and aluminum

  1. FDA Developments: Food Code 2013 and Proposed Trans Fat Determination

    NARCIS (Netherlands)

    Grossman, M.R.

    2014-01-01

    268 Reports EFFL 4|2014 USA FDA Developments: Food Code 2013 and Proposed Trans Fat Determination Margaret Rosso Grossman* I. Food Code 2013 and Food Code Reference System Since 1993, the US Food and Drug Administration has published a Food Code, now updated every four years. In November 2013, the

  2. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

    Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

    1991-01-01

    An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

  3. Interrelations of codes in human semiotic systems.

    OpenAIRE

    Somov, Georgij

    2016-01-01

    Codes can be viewed as mechanisms that enable relations of signs and their components, i.e., semiosis is actualized. The combinations of these relations produce new relations as new codes are building over other codes. Structures appear in the mechanisms of codes. Hence, codes can be described as transformations of structures from some material systems into others. Structures belong to different carriers, but exist in codes in their "pure" form. Building of codes over other codes fosters t...

  4. ETF system code: composition and applications

    International Nuclear Information System (INIS)

    Reid, R.L.; Wu, K.F.

    1980-01-01

    A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies, such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system

  5. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

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

  6. Development of the next generation code system as an engineering modelling language. 3. Study with prototyping. 2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Chiba, Go; Kasahara, Naoto; Ishikawa, Makoto

    2004-04-01

    In the fast reactor development, numerical simulations using analysis code play and important role for complementing theory and experiment. In order to efficiently advance the research and development of fast reactors, JNC promotes the development of next generation simulation code (NGSC). In this report, an investigation research result by prototyping which carried out for the conceptual design of the NGSC is described. From the viewpoint of the cooperative research with CEA (Commissariat a l'Energie Atomique) in France, a trend survey on several platforms for numerical analysis and an applicability evaluation of the SALOME platform in CEA for the NGSC were carried out. As a result of the evaluation, it is confirmed that the SALOME had been satisfied the features of efficiency, openness, universality, expansibility and completeness that are required by the NGSC. In addition, it is confirmed that the SALOME had the concept of the control layer required by the NGSC and would be one of the important candidates as a platform of the NGSC. In the field of the structure analysis, the prototype of the PRTS.NET code was reexamined from the viewpoint of class structure and input/output specification in order to improve the data processing efficiency and maintainability. In the field of the reactor physics analysis, a development test of a new code with C++ and a reuse test of an existing code written in Fortran was carried out in view of utilizing the SALOME for the NGSC. (author)

  7. Development of assessment system for tank earthquake-proof design (ASTEP code) installing automatic operation and knowledge database

    International Nuclear Information System (INIS)

    Maekawa, Akira; Suzuki, Michiaki; Fujii, Yuzo

    2004-01-01

    In a nuclear power station, seismic-proof design of the various tanks classified as auxiliary installation are required to follow technical guideline for the seismic-proof design of nuclear power station, which is called JEAC4601 for short in below. This guideline uses simple mechanical multi-mass model but a rather complicated evaluation method requires designers to have knowledge and experience and consumes both time and labor. On purpose to resolve those difficulties, Assessment System for Tank Earthquake-Proof Design, which is called ASTEP in short, has been developed and equipped with automated process and knowledge database. For this system, the targeted types of tank are a vertical cylindrical tank that has four supports or a skirt support, a horizontal cylindrical tank that has two saddle supports, and vertical cylindrical tank or water storage tank with a flat bottom. The system integrated all the seismic-proof design evaluation related tools and equipped with step by step menus in order of the flowchart, so enables designers to use them easily. In addition, it has a input aid that enables users to input with ease and a tool that automatically calculates input parameters. So this system reduces seismic-proof design evaluation related work load dramatically and also does not require much knowledge and experience related to this field. Further more, this system organized seismic-proof design related past statement and technical documents as a knowledge database so user could obtain the identical output as of the manual calculation results. Comparing output of ASTEP code and the manual calculation results of a typical tank that requires government approval of its design evaluation document, the error was within less than a percent so validity of the system was confirmed. This system has gained favorable comment during the trial run, and it was beyond our expectation. (author)

  8. Development of Evaluation Code for MUF Uncertainty

    International Nuclear Information System (INIS)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan

    2015-01-01

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities

  9. Development of Evaluation Code for MUF Uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities.

  10. Development and validation of sodium fire codes

    International Nuclear Information System (INIS)

    Morii, Tadashi; Himeno Yoshiaki; Miyake, Osamu

    1989-01-01

    Development, verification, and validation of the spray fire code, SPRAY-3M, the pool fire codes, SOFIRE-M2 and SPM, the aerosol behavior code, ABC-INTG, and the simultaneous spray and pool fires code, ASSCOPS, are presented. In addition, the state-of-the-art of development of the multi-dimensional natural convection code, SOLFAS, for the analysis of heat-mass transfer during a fire, is presented. (author)

  11. Reactor safety computer code development at INEL

    International Nuclear Information System (INIS)

    Johnsen, G.W.

    1985-01-01

    This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

  12. Development of MCNP interface code in HFETR

    International Nuclear Information System (INIS)

    Qiu Liqing; Fu Rong; Deng Caiyu

    2007-01-01

    In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)

  13. Implementing a modular system of computer codes

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-07-01

    A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out

  14. 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)

  15. Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Ama, T.; Palmiotti, G.; Taiwo, T.A.; Yang, W.S.

    2000-01-01

    The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects

  16. 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)

  17. Monte Carlo code development in Los Alamos

    International Nuclear Information System (INIS)

    Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.

    1974-01-01

    The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)

  18. Computer access security code system

    Science.gov (United States)

    Collins, Earl R., Jr. (Inventor)

    1990-01-01

    A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.

  19. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 2. Development and application of analytical evaluation system for thermal striping phenomena

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu

    2001-01-01

    Fluid-structure thermal interaction phenomena characterized by stationary random temperature fluctuations, namely thermal striping are observed in the downstream region such as a T-junction piping system of liquid metal fast reactors (LMFRs). Therefore, the piping wall located in the downstream region must be protected against the stationary random thermal process, which might induce high-cycle fatigue. This paper describes the evaluation system based on numerical simulation methods consisting of three thermohydraulics computer programs AQUA, DINUS-3 and THEMIS and of three thermomechanical computer programs BEMSET, FINAS and CANIS, for the thermal striping developed at Japan Nuclear Cycle Development Institute (JNC). Verification results for each computer code and the system are also introduced based on out-of-pile experimental data using water and sodium as working fluids. (author)

  20. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1979-03-01

    A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)

  1. Automatic Generation of Agents using Reusable Soft Computing Code Libraries to develop Multi Agent System for Healthcare

    OpenAIRE

    Priti Srinivas Sajja

    2015-01-01

    This paper illustrates architecture for a multi agent system in healthcare domain. The architecture is generic and designed in form of multiple layers. One of the layers of the architecture contains many proactive, co-operative and intelligent agents such as resource management agent, query agent, pattern detection agent and patient management agent. Another layer of the architecture is a collection of libraries to auto-generate code for agents using soft computing techni...

  2. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  3. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

  4. Next generation Zero-Code control system UI

    CERN Multimedia

    CERN. Geneva

    2017-01-01

    Developing ergonomic user interfaces for control systems is challenging, especially during machine upgrade and commissioning where several small changes may suddenly be required. Zero-code systems, such as *Inspector*, provide agile features for creating and maintaining control system interfaces. More so, these next generation Zero-code systems bring simplicity and uniformity and brake the boundaries between Users and Developers. In this talk we present *Inspector*, a CERN made Zero-code application development system, and we introduce the major differences and advantages of using Zero-code control systems to develop operational UI.

  5. The EGS5 Code System

    Energy Technology Data Exchange (ETDEWEB)

    Hirayama, Hideo; Namito, Yoshihito; /KEK, Tsukuba; Bielajew, Alex F.; Wilderman, Scott J.; U., Michigan; Nelson, Walter R.; /SLAC

    2005-12-20

    In the nineteen years since EGS4 was released, it has been used in a wide variety of applications, particularly in medical physics, radiation measurement studies, and industrial development. Every new user and every new application bring new challenges for Monte Carlo code designers, and code refinements and bug fixes eventually result in a code that becomes difficult to maintain. Several of the code modifications represented significant advances in electron and photon transport physics, and required a more substantial invocation than code patching. Moreover, the arcane MORTRAN3[48] computer language of EGS4, was highest on the complaint list of the users of EGS4. The size of the EGS4 user base is difficult to measure, as there never existed a formal user registration process. However, some idea of the numbers may be gleaned from the number of EGS4 manuals that were produced and distributed at SLAC: almost three thousand. Consequently, the EGS5 project was undertaken. It was decided to employ the FORTRAN 77 compiler, yet include as much as possible, the structural beauty and power of MORTRAN3. This report consists of four chapters and several appendices. Chapter 1 is an introduction to EGS5 and to this report in general. We suggest that you read it. Chapter 2 is a major update of similar chapters in the old EGS4 report[126] (SLAC-265) and the old EGS3 report[61] (SLAC-210), in which all the details of the old physics (i.e., models which were carried over from EGS4) and the new physics are gathered together. The descriptions of the new physics are extensive, and not for the faint of heart. Detailed knowledge of the contents of Chapter 2 is not essential in order to use EGS, but sophisticated users should be aware of its contents. In particular, details of the restrictions on the range of applicability of EGS are dispersed throughout the chapter. First-time users of EGS should skip Chapter 2 and come back to it later if necessary. With the release of the EGS4 version

  6. 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

  7. Development and application of the waste code

    International Nuclear Information System (INIS)

    Morison, I.W.

    1984-01-01

    This paper discusses the objectives and general approach underlying the Australian Code of Practice on the Management of Radioactive Wastes arising from the Mining and Milling of Radioactive Ores 1982. Background to the development of the Code is provided and the guidelines which supplement the Code are considered

  8. Development of a system code with CFD capability for analyzing turbulent mixed convection in gas-cooled reactors

    International Nuclear Information System (INIS)

    Kim, Hyeon Il

    2010-02-01

    In order to demonstrate the accuracy of predictions in a turbulent mixed convection regime in which both inertia and buoyancy force compete with each other, we found out that assessments done using a single-dimensional system code with a recently updated heat transfer package have shown that this approach cannot give a reasonable prediction of the wall temperature in a case involving strong heating, where the regime falls into turbulent mixed convection regime. It has been known that the main reason of this deficiency comes from the degraded heat transfer in turbulent mixed convection regime, which is below that of convective heat transfer during turbulent forced convection. We investigated two mechanisms that cause this deterioration in convective heat transfer influenced by buoyancy: (1) modification of turbulence, also known as the direct (structural) effect, through the buoyancy-induced production of turbulent kinetic energy: and (2) an indirect (external) effect that occurs through modification of the mean flow. We investigated the Launder-Sharma model of turbulence whether it can appropriately represent the mechanisms causing the degraded heat transfer in Computational Fluid Dynamics (CFD). We found out that this model can capture low Re effects such that a non-equilibrium turbulent boundary layer in turbulent mixed convection regime can be resolved. The model was verified and validated extensively initially with the commercial CFD code, Fluent with a user application package known as the User Defined Function (UDF). The results from this implementation were compared to a set of data that included (1) an experimental data commonly accepted as a standardized problem to verify a turbulent flow, (2) the results from a Direct Numerical Simulation (DNS) in a turbulent forced and mixed convection regime, (3) empirical correlations regarding the friction coefficient and the non-dimensional heat transfer coefficient, the Nusselt number for a turbulent forced

  9. Development of a computer code system for selecting off-site protective action in radiological accidents based on the multiobjective optimization method

    International Nuclear Information System (INIS)

    Ishigami, Tsutomu; Oyama, Kazuo

    1989-09-01

    This report presents a new method to support selection of off-site protective action in nuclear reactor accidents, and provides a user's manual of a computer code system, PRASMA, developed using the method. The PRASMA code system gives several candidates of protective action zones of evacuation, sheltering and no action based on the multiobjective optimization method, which requires objective functions and decision variables. We have assigned population risks of fatality, injury and cost as the objective functions, and distance from a nuclear power plant characterizing the above three protective action zones as the decision variables. (author)

  10. Plotting system for the MINCS code

    International Nuclear Information System (INIS)

    Watanabe, Tadashi

    1990-08-01

    The plotting system for the MINCS code is described. The transient two-phase flow analysis code MINCS has been developed to provide a computational tool for analysing various two-phase flow phenomena in one-dimensional ducts. Two plotting systems, namely the SPLPLOT system and the SDPLOT system, can be used as the plotting functions. The SPLPLOT system is used for plotting time transients of variables, while the SDPLOT system is for spatial distributions. The SPLPLOT system is based on the SPLPACK system, which is used as a general tool for plotting results of transient analysis codes or experiments. The SDPLOT is based on the GPLP program, which is also regarded as one of the general plotting programs. In the SPLPLOT and the SDPLOT systems, the standardized data format called the SPL format is used in reading data to be plotted. The output data format of MINCS is translated into the SPL format by using the conversion system called the MINTOSPL system. In this report, how to use the plotting functions is described. (author)

  11. Development of system based code for integrity of FBR. Fundamental probabilistic approach, Part 1: Model calculation of creep-fatigue damage (Research report)

    International Nuclear Information System (INIS)

    Kawasaki, Nobuchika; Asayama, Tai

    2001-09-01

    Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow, a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks. (author)

  12. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  13. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  14. SWAAM code development, verification and application to steam generator design

    International Nuclear Information System (INIS)

    Shin, Y.W.; Valentin, R.A.

    1990-01-01

    This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs

  15. Expansion of the CHR bone code system

    International Nuclear Information System (INIS)

    Farnham, J.E.; Schlenker, R.A.

    1976-01-01

    This report describes the coding system used in the Center for Human Radiobiology (CHR) to identify individual bones and portions of bones of a complete skeletal system. It includes illustrations of various bones and bone segments with their respective code numbers. Codes are also presented for bone groups and for nonbone materials

  16. Development of the CELVA-1D code to evaluate the safety of an air-ventilation system during postulated fire and explosion in the reprocessing plant. Contract research

    International Nuclear Information System (INIS)

    Nishio, Gunji; Watanabe, Kouji; Kouno, Kouji; Yamazaki, Noboru; Mukaide, Shigeo; Yoshioka, Itsuo

    1998-03-01

    The CELVA-1D computer code was developed to evaluate the confinement of radioactive materials during postulated fire and explosion in a cell of nuclear fuel reprocessing plants. The CELVA-1D code calculates a response of temperature, pressure, flow velocity of fluid in an air-ventilation system of the plants by one-dimensional thermofluid analysis and calculates an ability to confine radioactive aerosol particles by transport, deposition, and HEPA filtration. The mathematical models in CELVA-1D were verified by comparison of the calculation with the result of JAERI's demonstration tests simulating hypothetical fire and explosion accidents in the cell. (author)

  17. Development of the CELVA-1D code to evaluate the safety of an air-ventilation system during postulated fire and explosion in the reprocessing plant. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Nishio, Gunji; Watanabe, Kouji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kouno, Kouji; Yamazaki, Noboru; Mukaide, Shigeo; Yoshioka, Itsuo

    1998-03-01

    The CELVA-1D computer code was developed to evaluate the confinement of radioactive materials during postulated fire and explosion in a cell of nuclear fuel reprocessing plants. The CELVA-1D code calculates a response of temperature, pressure, flow velocity of fluid in an air-ventilation system of the plants by one-dimensional thermofluid analysis and calculates an ability to confine radioactive aerosol particles by transport, deposition, and HEPA filtration. The mathematical models in CELVA-1D were verified by comparison of the calculation with the result of JAERI`s demonstration tests simulating hypothetical fire and explosion accidents in the cell. (author)

  18. Development of a Wrapper Object for MARS TH Systems Code and Its Applications in Object Oriented Programs

    International Nuclear Information System (INIS)

    Park, Sun Byung; Lee, Young Jin; Kim, Hyong Chol; Han, Sam Hee; Kim, Hyun Jik

    2013-01-01

    TMARS is written for the object pascal program language, and 'wraps' the Dynamic Link Library (DLL) manifestation of the MARS-KS code written in Fortran 90. TMARS behaves as a true object and it can be instantiated, inherited, and its methods overloaded. The functionality of TMARS was verified and demonstrated using two programs built under object oriented program environment. One is a text based program for reviewing the data interface of TMARS, and the other is a graphic intensive prototype NPA program for testing the overall performance of TMARS. The prototype NPA was also used to assess the real-time capability of TMARS. The demonstration programs show that application of TMARS is straight forward and that its functions facilitate easy application developments. TMARS, a wrapper object class encapsulating the calculation functions of MARS-KS code, was successfully developed and verification of its functionality was carried out using custom made programs. The verification results show that TMARS is capable of providing reliable TH calculation results and sufficient performance to realize real time calculations

  19. RDS; A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

    2013-01-01

    Full-text: Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBIMOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges. (author)

  20. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat

    2014-01-01

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges

  1. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  2. Coding-Spreading Tradeoff in CDMA Systems

    National Research Council Canada - National Science Library

    Bolas, Eduardo

    2002-01-01

    .... Comparing different combinations of coding and spreading with a traditional DS-CDMA, as defined in the IS-95 standard, allows the criteria to be defined for the best coding-spreading tradeoff in CDMA systems...

  3. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Tobias

    2013-07-03

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  4. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    International Nuclear Information System (INIS)

    Hartmann, Tobias

    2013-01-01

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  5. Code development for nuclear reactor simulation

    International Nuclear Information System (INIS)

    Chauliac, C.; Verwaerde, D.; Pavageau, O.

    2006-01-01

    Full text of publication follows: Since several years, CEA, EDF and FANP have developed several numerical codes which are currently used for nuclear industry applications and will be remain in use for the coming years. Complementary to this set of codes and in order to better meet the present and future needs, a new system is being developed through a joint venture between CEA, EDF and FANP, with a ten year prospect and strong intermediate milestones. The focus is put on a multi-scale and multi-physics approach enabling to take into account phenomena from microscopic to macroscopic scale, and to describe interactions between various physical fields such as neutronics (DESCARTES), thermal-hydraulics (NEPTUNE) and fuel behaviour (PLEIADES). This approach is based on a more rational design of the softwares and uses a common integration platform providing pre-processing, supervision of computation and post-processing. This paper will describe the overall system under development and present the first results obtained. (authors)

  6. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  7. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  8. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)

  9. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  10. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  11. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  12. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  13. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  14. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  15. Development of Probabilistic Internal Dosimetry Computer Code

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Siwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Tae-Eun [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Lee, Jai-Ki [Korean Association for Radiation Protection, Seoul (Korea, Republic of)

    2017-02-15

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5{sup th}, 5{sup th}, median, 95{sup th}, and 97.5{sup th} percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various

  16. Development of Probabilistic Internal Dosimetry Computer Code

    International Nuclear Information System (INIS)

    Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki

    2017-01-01

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5 th , 5 th , median, 95 th , and 97.5 th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases

  17. High dynamic range coding imaging system

    Science.gov (United States)

    Wu, Renfan; Huang, Yifan; Hou, Guangqi

    2014-10-01

    We present a high dynamic range (HDR) imaging system design scheme based on coded aperture technique. This scheme can help us obtain HDR images which have extended depth of field. We adopt Sparse coding algorithm to design coded patterns. Then we utilize the sensor unit to acquire coded images under different exposure settings. With the guide of the multiple exposure parameters, a series of low dynamic range (LDR) coded images are reconstructed. We use some existing algorithms to fuse and display a HDR image by those LDR images. We build an optical simulation model and get some simulation images to verify the novel system.

  18. MIDAS/PK code development using point kinetics model

    International Nuclear Information System (INIS)

    Song, Y. M.; Park, S. H.

    1999-01-01

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

  19. NASA space radiation transport code development consortium

    International Nuclear Information System (INIS)

    Townsend, L. W.

    2005-01-01

    Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)

  20. Computer-assisted Particle-in-Cell code development

    International Nuclear Information System (INIS)

    Kawata, S.; Boonmee, C.; Teramoto, T.; Drska, L.; Limpouch, J.; Liska, R.; Sinor, M.

    1997-12-01

    This report presents a new approach for an electromagnetic Particle-in-Cell (PIC) code development by a computer: in general PIC codes have a common structure, and consist of a particle pusher, a field solver, charge and current density collections, and a field interpolation. Because of the common feature, the main part of the PIC code can be mechanically developed on a computer. In this report we use the packages FIDE and GENTRAN of the REDUCE computer algebra system for discretizations of field equations and a particle equation, and for an automatic generation of Fortran codes. The approach proposed is successfully applied to the development of 1.5-dimensional PIC code. By using the generated PIC code the Weibel instability in a plasma is simulated. The obtained growth rate agrees well with the theoretical value. (author)

  1. Development of the system based code. v. 5. Method of margin exchange. pt. 2. Determination of quality assurance index based on a 'Vector Method'

    International Nuclear Information System (INIS)

    Asayama, Tai

    2003-03-01

    For the commercialization of fast breeder reactors, 'System Based Code', a completely new scheme of a code on structural integrity, is being developed. One of the distinguished features of the System Based Code is that it is able to determine a reasonable total margin on a structural of system, by allowing the exchanges of margins between various technical items. Detailed estimation of failure probability of a given combination of technical items and its comparison with a target value is one way to achieve this. However, simpler and easier methods that allow margin exchange without detailed calculation of failure probability are desirable in design. The authors have developed a simplified method such as a 'design factor method' from this viewpoint. This report describes a 'Vector Method', which was been newly developed. Following points are reported: 1) The Vector Method allows margin exchange evaluation on an 'equi-quality assurance plane' using vector calculation. Evaluation is easy and sufficient accuracy is achieved. The equi-quality assurance plane is obtained by a projection of an 'equi-failure probability surface in a n-dimensional space, which is calculated beforehand for typical combinations of design variables. 2) The Vector Method is considered to give the 'Quality Assurance Index Method' a probabilistic interpretation. 3) An algebraic method was proposed for the calculation of failure probabilities, which is necessary to obtain a equi-failure probability surface. This method calculates failure probabilities without using numerical methods such as Monte Carlo simulation or numerical integration. Under limited conditions, this method is quite effective compared to numerical methods. 4) An illustration of the procedure of margin exchange evaluation is given. It may be possible to use this method to optimize ISI plans; even it is not fully implemented in the System Based Code. (author)

  2. Development of an automatic validation system for simulation codes of the fusion research; Entwicklung eines automatischen Validierungssystems fuer Simulationscodes der Fusionsforschung

    Energy Technology Data Exchange (ETDEWEB)

    Galonska, Andreas

    2010-03-15

    In the present master thesis the development oa an automatic validation system for the simulation code ERO is documented. This 3D Monte-carlo code models the transport of impurities as well as plasma-wall interaction processes and has great importance for the fusion research. The validation system is based on JuBE (Julich Benchmarking Environment), the flexibility of which allows a slight extension of the system to other codes, for instance such, which are operated in the framework of the EU Task Force ITM (Integrated Tokamak Modelling). The chosen solution - JuBE and a special program for the ''intellectual'' comparison of actual and reference-edition data of ERO is described and founded. The use of this program and the configuration of JuBE are detailedly described. Simulations to different plasma experiments, which serve as reference cases for the automatic validation, are explained. The working of the system is illustrated by the description of a test case. This treats the failure localization and improvement in the parallelization of an important ERO module (tracking of physically eroded particle). It is demonstrated, how the system reacts in an erroneous validation and the subsequently performed error correction leads to a positive result. Finally a speed-up curve of the parallelization is established by means of the output data of JuBE.

  3. CATHARE code development and assessment methodologies

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. Development and Verification of Behavior of Tritium Analytic Code (BOTANIC)

    International Nuclear Information System (INIS)

    Park, Min Young; Kim, Eung Soo

    2014-01-01

    VHTR, one of the Generation IV reactor concepts, has a relatively high operation temperature and is usually suggested as a heat source for many industrial processes, including hydrogen production process. Thus, it is vital to trace tritium behavior in the VHTR system and the potential permeation rate to the industrial process. In other words, tritium is a crucial issue in terms of safety in the fission reactor system. Therefore, it is necessary to understand the behavior of tritium and the development of the tool to enable this is vital.. In this study, a Behavior of Tritium Analytic Code (BOTANIC) an analytic tool which is capable of analyzing tritium behavior is developed using a chemical process code called gPROMS. BOTANIC was then further verified using the analytic solutions and benchmark codes such as Tritium Permeation Analysis Code (TPAC) and COMSOL. In this study, the Behavior of Tritium Analytic Code, BOTANIC, has been developed using a chemical process code called gPROMS. The code has several distinctive features including non-diluted assumption, flexible applications and adoption of distributed permeation model. Due to these features, BOTANIC has the capability to analyze a wide range of tritium level systems and has a higher accuracy as it has the capacity to solve distributed models. BOTANIC was successfully developed and verified using analytical solution and the benchmark code calculation result. The results showed very good agreement with the analytical solutions and the calculation results of TPAC and COMSOL. Future work will be focused on the total system verification

  5. SRAC95; general purpose neutronics code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Tsuchihashi, Keichiro; Kaneko, Kunio.

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author)

  6. SRAC95; general purpose neutronics code system

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).

  7. Development of ADINA-J-integral code

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1988-07-01

    A general purpose finite element program ADINA (Automatic Dynamic Incremental Nonlinear Analysis), which was developed by Bathe et al., was revised to be able to calculate the J- and J-integral. This report introduced the numerical method to add this capability to the code, and the evaluation of the revised ADINA-J code by using a few of examples of the J estimation model, i.e. a compact tension specimen, a center cracked panel subjected to dynamic load, and a thick shell cylinder having inner axial crack subjected to thermal load. The evaluation testified the function of the revised code. (author)

  8. Discrete fracture network code development

    Energy Technology Data Exchange (ETDEWEB)

    Dershowitz, W.; Doe, T.; Shuttle, D.; Eiben, T.; Fox, A.; Emsley, S.; Ahlstrom, E. [Golder Associates Inc., Redmond, Washington (United States)

    1999-02-01

    This report presents the results of fracture flow model development and application performed by Golder Associates Inc. during the fiscal year 1998. The primary objective of the Golder Associates work scope was to provide theoretical and modelling support to the JNC performance assessment effort in fiscal year 2000. In addition, Golder Associates provided technical support to JNC for the Aespoe project. Major efforts for performance assessment support included extensive flow and transport simulations, analysis of pathway simplification, research on excavation damage zone effects, software verification and cross-verification, and analysis of confidence bounds on Monte Carlo simulations. In addition, a Fickian diffusion algorithm was implemented for Laplace Transform Galerkin solute transport. Support for the Aespoe project included predictive modelling of sorbing tracer transport in the TRUE-1 rock block, analysis of 1 km geochemical transport pathways for Task 5', and data analysis and experimental design for the TRUE Block Scale experiment. Technical information about Golder Associates support to JNC is provided in the appendices to this report. (author)

  9. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  10. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  12. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  13. COSINE software development based on code generation technology

    International Nuclear Information System (INIS)

    Ren Hao; Mo Wentao; Liu Shuo; Zhao Guang

    2013-01-01

    The code generation technology can significantly improve the quality and productivity of software development and reduce software development risk. At present, the code generator is usually based on UML model-driven technology, which can not satisfy the development demand of nuclear power calculation software. The feature of scientific computing program was analyzed and the FORTRAN code generator (FCG) based on C# was developed in this paper. FCG can generate module variable definition FORTRAN code automatically according to input metadata. FCG also can generate memory allocation interface for dynamic variables as well as data access interface. FCG was applied to the core and system integrated engine for design and analysis (COSINE) software development. The result shows that FCG can greatly improve the development efficiency of nuclear power calculation software, and reduce the defect rate of software development. (authors)

  14. Los Alamos radiation transport code system on desktop computing platforms

    International Nuclear Information System (INIS)

    Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

    1990-01-01

    The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

  15. Development of 2-d cfd code

    International Nuclear Information System (INIS)

    Mirza, S.A.

    1999-01-01

    In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)

  16. Health Code Number (HCN) Development Procedure

    Energy Technology Data Exchange (ETDEWEB)

    Petrocchi, Rocky; Craig, Douglas K.; Bond, Jayne-Anne; Trott, Donna M.; Yu, Xiao-Ying

    2013-09-01

    This report provides the detailed description of health code numbers (HCNs) and the procedure of how each HCN is assigned. It contains many guidelines and rationales of HCNs. HCNs are used in the chemical mixture methodology (CMM), a method recommended by the department of energy (DOE) for assessing health effects as a result of exposures to airborne aerosols in an emergency. The procedure is a useful tool for proficient HCN code developers. Intense training and quality assurance with qualified HCN developers are required before an individual comprehends the procedure to develop HCNs for DOE.

  17. Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

    International Nuclear Information System (INIS)

    Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

    2004-01-01

    Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials

  18. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  19. Development of a Performance Analysis Code for the Off-design conditions of a S-CO2 Brayton Cycle Energy Conversion System

    International Nuclear Information System (INIS)

    Yoo, Yong-Hwan; Cha, Jae-Eun; Lee, Tae-Ho; Eoh, Jae-Hyuk; Kim, Seong-O

    2008-01-01

    For the development of a supercritical carbon dioxide (S-CO2) Brayton cycle energy conversion system coupled to KALIMER-600, a thermal balance has been established on 100% power operating conditions including all the reactor system models such as a primary heat transport system (PHTS), an intermediate heat transport system (IHTS), and an energy conversion system. The S-CO2 Brayton cycle energy conversion system consists of a sodium-CO2 heat exchanger (Hx), turbine, high temperature recuperate (HTR), low temperature recuperate (LTR), precooler, compressor no.1, and compressor no.2. Two compressors were employed to avoid a sharp change of the physical properties near their critical point with a corresponding pressure. The component locations and their operating conditions are illustrated. Energy balance of the power conversion system in KALIMER-600 was designed with the full power condition of each component. Therefore, to predict the off-design conditions and to evaluate each component, an off-design performance analysis code should be accomplished. An off-design performance analysis could be classified into overall system control logic and local system control logic. The former means that mass flow rate and power are controlled by valves, and the latter implies that a bypass or inventory control is an admitted system balance. The ultimate goal of this study is development of the overall system control logic

  20. 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.)

  1. Development of a computational system for radiotherapic planning with the IMRT technique applied to the MCNP computer code with 3D graphic interface for voxel models

    International Nuclear Information System (INIS)

    Fonseca, Telma Cristina Ferreira

    2009-01-01

    The Intensity Modulated Radiation Therapy - IMRT is an advanced treatment technique used worldwide in oncology medicine branch. On this master proposal was developed a software package for simulating the IMRT protocol, namely SOFT-RT which attachment the research group 'Nucleo de Radiacoes Ionizantes' - NRI at UFMG. The computational system SOFT-RT allows producing the absorbed dose simulation of the radiotherapic treatment through a three-dimensional voxel model of the patient. The SISCODES code, from NRI, research group, helps in producing the voxel model of the interest region from a set of CT or MRI digitalized images. The SOFT-RT allows also the rotation and translation of the model about the coordinate system axis for better visualization of the model and the beam. The SOFT-RT collects and exports the necessary parameters to MCNP code which will carry out the nuclear radiation transport towards the tumor and adjacent healthy tissues for each orientation and position of the beam planning. Through three-dimensional visualization of voxel model of a patient, it is possible to focus on a tumoral region preserving the whole tissues around them. It takes in account where exactly the radiation beam passes through, which tissues are affected and how much dose is applied in both tissues. The Out-module from SOFT-RT imports the results and express the dose response superimposing dose and voxel model in gray scale in a three-dimensional graphic representation. The present master thesis presents the new computational system of radiotherapic treatment - SOFT-RT code which has been developed using the robust and multi-platform C ++ programming language with the OpenGL graphics packages. The Linux operational system was adopted with the goal of running it in an open source platform and free access. Preliminary simulation results for a cerebral tumor case will be reported as well as some dosimetric evaluations. (author)

  2. Development of code PRETOR for stellarator simulation

    International Nuclear Information System (INIS)

    Dies, J.; Fontanet, J.; Fontdecaba, J.M.; Castejon, F.; Alejandre, C.

    1998-01-01

    The Department de Fisica i Enginyeria Nuclear (DFEN) of the UPC has some experience in the development of the transport code PRETOR. This code has been validated with shots of DIII-D, JET and TFTR, it has also been used in the simulation of operational scenarios of ITER fast burnt termination. Recently, the association EURATOM-CIEMAT has started the operation of the TJ-II stellarator. Due to the need of validating the results given by others transport codes applied to stellarators and because all of them made some approximations, as a averaging magnitudes in each magnetic surface, it was thought suitable to adapt the PRETOR code to devices without axial symmetry, like stellarators, which is very suitable for the specific needs of the study of TJ-II. Several modifications are required in PRETOR; the main concerns to the models of: magnetic equilibrium, geometry and transport of energy and particles. In order to solve the complex magnetic equilibrium geometry the powerful numerical code VMEC has been used. This code gives the magnetic surface shape as a Fourier series in terms of the harmonics (m,n). Most of the geometric magnitudes are also obtained from the VMEC results file. The energy and particle transport models will be replaced by other phenomenological models that are better adapted to stellarator simulation. Using the proposed models, it is pretended to reproduce experimental data available from present stellarators, given especial attention to the TJ-II of the association EURATOM-CIEMAT. (Author)

  3. Coupled geochemical and solute transport code development

    International Nuclear Information System (INIS)

    Morrey, J.R.; Hostetler, C.J.

    1985-01-01

    A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

  4. Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR

    International Nuclear Information System (INIS)

    Chaudri, Khurrum Saleem; Su Yali; Chen Ronghua; Tian Wenxi; Su Guanghui; Qiu Suizheng

    2012-01-01

    Highlights: ► A tool is developed for coupled neutronics/thermal-hydraulic analysis for SCWR. ► For thermal hydraulic analysis, a sub-channel code SACoS is developed and verified. ► Coupled analysis agree quite well with the reference calculations. ► Different choice of important parameters makes huge difference in design calculations. - Abstract: Supercritical Water Reactor (SCWR) is one of the promising reactors from the list of fourth generation of nuclear reactors. High thermal efficiency and low cost of electricity make it an attractive option in the era of growing energy demand. An almost seven fold density variation for coolant/moderator along the active height does not allow the use of constant density assumption for design calculations, as used for previous generations of reactors. The advancement in computer technology gives us the superior option of performing coupled analysis. Thermal hydraulics calculations of supercritical water systems present extra challenges as not many computational tools are available to perform that job. This paper introduces a new sub-channel code called Sub-channel Analysis Code of SCWR (SACoS) and its application in coupled analyses of High Performance Light Water Reactor (HPLWR). SACoS can compute the basic thermal hydraulic parameters needed for design studies of a supercritical water reactor. Multiple heat transfer and pressure drop correlations are incorporated in the code according to the flow regime. It has the additional capability of calculating the thermal hydraulic parameters of moderator flowing in water box and between fuel assemblies under co-current or counter current flow conditions. Using MCNP4c and SACoS, a coupled system has been developed for SCWR design analyses. The developed coupled system is verified by performing and comparing HPLWR calculations. The results were found to be in very good agreement. Significant difference between the results was seen when Doppler feedback effect was included in

  5. 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

  6. Graphical user interface development for the MARS code

    International Nuclear Information System (INIS)

    Jeong, J.-J.; Hwang, M.; Lee, Y.J.; Kim, K.D.; Chung, B.D.

    2003-01-01

    KAERI has developed the best-estimate thermal-hydraulic system code MARS using the RELAP5/MOD3 and COBRA-TF codes. To exploit the excellent features of the two codes, we consolidated the two codes. Then, to improve the readability, maintainability, and portability of the consolidated code, all the subroutines were completely restructured by employing a modular data structure. At present, a major part of the MARS code development program is underway to improve the existing capabilities. The code couplings with three-dimensional neutron kinetics, containment analysis, and transient critical heat flux calculations have also been carried out. At the same time, graphical user interface (GUI) tools have been developed for user friendliness. This paper presents the main features of the MARS GUI. The primary objective of the GUI development was to provide a valuable aid for all levels of MARS users in their output interpretation and interactive controls. Especially, an interactive control function was designed to allow operator actions during simulation so that users can utilize the MARS code like conventional nuclear plant analyzers (NPAs). (author)

  7. 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)

  8. 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

  9. 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)

  10. Developing an Australian code of construction ethics

    Directory of Open Access Journals (Sweden)

    Sean Francis McCarthy

    2012-05-01

    Full Text Available This article looks at the increasing need to consider the role of ethics in construction. The industry, historically, has been challenged by allegations of a serious shortfall in ethical standards. Only limited attempts to date in Australia have been made to address that concern. Any ethical analysis should consider the definition of ethics and its historical development. This paper considers major historical developments in ethical thinking as well as contemporary thinking on ethics for professional sub-sets. A code could be developed specific to construction. Current methods of addressing ethics in construction and in other industries are also reviewed. This paper argues that developing a code of ethics, supported by other measures is the way forward. The author’s aim is to promote further discussion and promote the drafting of a code. This paper includes a summary of other ethical codes that may provide a starting point. The time for reform is upon us, and there is an urgent need for an independent body to take the lead, for fear of floundering and having only found ‘another debating topic’ (Uff 2006.

  11. A bar coding system for environmental projects

    International Nuclear Information System (INIS)

    Barber, R.B.; Hunt, B.J.; Burgess, G.M.

    1988-01-01

    This paper presents BeCode systems, a bar coding system which provides both nuclear and commercial clients with a data capture and custody management program that is accurate, timely, and beneficial to all levels of project operations. Using bar code identifiers is an essentially paperless and error-free method which provides more efficient delivery of data through its menu card-driven structure, which speeds collection of essential data for uploading to a compatible device. The effects of this sequence include real-time information for operator analysis, management review, audits, planning, scheduling, and cost control

  12. Arabic Natural Language Processing System Code Library

    Science.gov (United States)

    2014-06-01

    Adelphi, MD 20783-1197 This technical note provides a brief description of a Java library for Arabic natural language processing ( NLP ) containing code...for training and applying the Arabic NLP system described in the paper "A Cross-Task Flexible Transition Model for Arabic Tokenization, Affix...and also English) natural language processing ( NLP ), containing code for training and applying the Arabic NLP system described in Stephen Tratz’s

  13. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  14. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  15. Development of a fuel depletion sensitivity calculation module for multi-cell problems in a deterministic reactor physics code system CBZ

    International Nuclear Information System (INIS)

    Chiba, Go; Kawamoto, Yosuke; Narabayashi, Tadashi

    2016-01-01

    Highlights: • A new functionality of fuel depletion sensitivity calculations is developed in a code system CBZ. • This is based on the generalized perturbation theory for fuel depletion problems. • The theory with a multi-layer depletion step division scheme is described. • Numerical techniques employed in actual implementation are also provided. - Abstract: A new functionality of fuel depletion sensitivity calculations is developed as one module in a deterministic reactor physics code system CBZ. This is based on the generalized perturbation theory for fuel depletion problems. The theory for fuel depletion problems with a multi-layer depletion step division scheme is described in detail. Numerical techniques employed in actual implementation are also provided. Verification calculations are carried out for a 3 × 3 multi-cell problem consisting of two different types of fuel pins. It is shown that the sensitivities of nuclide number densities after fuel depletion with respect to the nuclear data calculated by the new module agree well with reference sensitivities calculated by direct numerical differentiation. To demonstrate the usefulness of the new module, fuel depletion sensitivities in different multi-cell arrangements are compared and non-negligible differences are observed. Nuclear data-induced uncertainties of nuclide number densities obtained with the calculated sensitivities are also compared.

  16. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

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

  17. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo

    1989-01-01

    Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)

  18. 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)

  19. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

    Ryman, J.C.; Cristy, M.; Eckerman, K.F.; Davis, J.L.; Tang, J.S.; Kerr, G.D.

    1986-01-01

    Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods

  20. 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)

  1. Software requirements specification document for the AREST code development

    International Nuclear Information System (INIS)

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

    1993-11-01

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

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

    International Nuclear Information System (INIS)

    1995-08-01

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

  3. A computerized energy systems code and information library at Soreq

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, I; Shapira, M; Caner, D; Sapier, D [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

    1996-12-01

    In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors).

  4. A computerized energy systems code and information library at Soreq

    International Nuclear Information System (INIS)

    Silverman, I.; Shapira, M.; Caner, D.; Sapier, D.

    1996-01-01

    In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors)

  5. Development of a Nuclear Steam Supply System Thermal-Hydraulic Module for the Nuclear Power Plant Simulator Using a Best-Estimate Code, RETRAN

    International Nuclear Information System (INIS)

    Suh, Jae Seung

    2004-08-01

    The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually used very simplified physical models for the real-time simulation of Ness thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, a realistic NSSS thermal-hydraulic program ARTS has been developed, it was based on the RETRAN code for the improvement of the Nuclear Power Plant full-scope simulator. Since ARTS is a generalized code to solve a simultaneous equation system, the smaller time-step size should be used if converged solution could not obtain even in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. The PRT(Pressurizer Relief Tank) is a good example, which requires a dedicated model. The PRT consists of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume may limit the time-step size if it is modeled with a general control volume. To mitigate the time-step size reduction due to convergence failure at this component using RETRAN, the PRT model was developed as a dedicated model. The dedicated model was expected to provide reasonable results without convergence problem in the analysis of the system transients. The ARTS code guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there are some possibilities of calculation failure in the

  6. Development of steam explosion simulation code JASMINE

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nagano, Katsuhiro; Araki, Kazuhiro

    1995-11-01

    A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author).

  7. Development of a tritium dispersion code

    International Nuclear Information System (INIS)

    Bell, R.P.; Davis, M.W.; Joseph, S.; Wong, K.Y.

    1985-01-01

    This paper describes the development and verification of a computer code designed to calculate the radiation dose to man following acute or chronic atmospheric releases of tritium gas and oxide from a point source. The Ontario Hydro Tritium Dispersion Code calculates tritium concentrations in air, soil, and vegetation and doses to man resulting from inhalation/immersion and ingestion of food, milk meat and water. The deposition of HT to soil, conversion of HT to HTO by soil enzymes and resuspension of HTO to air have been incorporated into the terrestrial compartment model and are unique features of the code. Sensitivity analysis has identified the HT deposition velocity and the equivalent water depth of the vegetation compartment as two parameters which have a strong influence on dose calculations. Tritium concentrations in vegetation and soil calculated by the code were in reasonable agreement with experimental results. The radiological significance of including the mechanisms of HT to HTO conversion and resuspension of HTO to air is illustrated

  8. Development of steam explosion simulation code JASMINE

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun; Nagano, Katsuhiro; Araki, Kazuhiro.

    1995-11-01

    A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author)

  9. Towards Product Lining Model-Driven Development Code Generators

    OpenAIRE

    Roth, Alexander; Rumpe, Bernhard

    2015-01-01

    A code generator systematically transforms compact models to detailed code. Today, code generation is regarded as an integral part of model-driven development (MDD). Despite its relevance, the development of code generators is an inherently complex task and common methodologies and architectures are lacking. Additionally, reuse and extension of existing code generators only exist on individual parts. A systematic development and reuse based on a code generator product line is still in its inf...

  10. 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

  11. Fast decoding algorithms for coded aperture systems

    International Nuclear Information System (INIS)

    Byard, Kevin

    2014-01-01

    Fast decoding algorithms are described for a number of established coded aperture systems. The fast decoding algorithms for all these systems offer significant reductions in the number of calculations required when reconstructing images formed by a coded aperture system and hence require less computation time to produce the images. The algorithms may therefore be of use in applications that require fast image reconstruction, such as near real-time nuclear medicine and location of hazardous radioactive spillage. Experimental tests confirm the efficacy of the fast decoding techniques

  12. A study on the nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Lee, Jong Bok; Choi, Young Gil; Suh, Soong Hyok; Kang, Byong Heon; Kim, Hee Kyung; Kim, Ko Ryeo; Park, Soo Jin

    1990-12-01

    According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)

  13. SCDAP/RELAP5/MOD3 code development

    International Nuclear Information System (INIS)

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-01-01

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

  14. SCDAP/RELAP5 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Hohorst, J.K.

    1996-01-01

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

  15. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs

  16. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Rollstin, J.A.; Chanin, D.I.; Jow, H.N.

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management

  17. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  18. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  19. SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi

    2003-01-01

    SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)

  20. Development of throughflow calculation code for axial flow compressors

    International Nuclear Information System (INIS)

    Kim, Ji Hwan; Kim, Hyeun Min; No, Hee Cheon

    2005-01-01

    The power conversion systems of the current HTGRs are based on closed Brayton cycle and major concern is thermodynamic performance of the axial flow helium gas turbines. Particularly, the helium compressor has some unique design challenges compared to the air-breathing compressor such as high hub-to-tip ratios throughout the machine and a large number of stages due to the physical property of the helium and thermodynamic cycle. Therefore, it is necessary to develop a design and analysis code for helium compressor that can estimate the design point and off-design performance accurately. KAIST nuclear system laboratory has developed a compressor design and analysis code by means of throughflow calculation and several loss models. This paper presents the outline of the development of a throughflow calculation code and its verification results

  1. Using Coding Apps to Support Literacy Instruction and Develop Coding Literacy

    Science.gov (United States)

    Hutchison, Amy; Nadolny, Larysa; Estapa, Anne

    2016-01-01

    In this article the authors present the concept of Coding Literacy and describe the ways in which coding apps can support the development of Coding Literacy and disciplinary and digital literacy skills. Through detailed examples, we describe how coding apps can be integrated into literacy instruction to support learning of the Common Core English…

  2. Development of Nuclear Energy Security Code

    International Nuclear Information System (INIS)

    Shimamura, Takehisa; Suzuki, Atsuyuki; Okubo, Hiroo; Kikuchi, Masahiro.

    1990-01-01

    In establishing of the nuclear fuel cycle in Japan that have a vulnerability in own energy structure, an effectiveness of energy security should be taken into account as well as an economy based on the balance of supply and demand of nuclear fuels. NMCC develops the 'Nuclear Energy Security Code' which was able to evaluate the effectiveness of energy security. Evaluation method adopted in this code is 'Import Premium' which was proposed in 'World Oil', EMF Report 6. The viewpoints of evaluation are as follows: 1. How much uranium fuel quantity can be reduced by using plutonium fuel? 2. How much a sudden rise of fuel cost can be absorbed by establishing the plutonium cycle beforehand the energy crisis? (author)

  3. Development of the point-depletion code DEPTH

    International Nuclear Information System (INIS)

    She, Ding; Wang, Kan; Yu, Ganglin

    2013-01-01

    Highlights: ► The DEPTH code has been developed for the large-scale depletion system. ► DEPTH uses the data library which is convenient to couple with MC codes. ► TTA and matrix exponential methods are implemented and compared. ► DEPTH is able to calculate integral quantities based on the matrix inverse. ► Code-to-code comparisons prove the accuracy and efficiency of DEPTH. -- Abstract: The burnup analysis is an important aspect in reactor physics, which is generally done by coupling of transport calculations and point-depletion calculations. DEPTH is a newly-developed point-depletion code of handling large burnup depletion systems and detailed depletion chains. For better coupling with Monte Carlo transport codes, DEPTH uses data libraries based on the combination of ORIGEN-2 and ORIGEN-S and allows users to assign problem-dependent libraries for each depletion step. DEPTH implements various algorithms of treating the stiff depletion systems, including the Transmutation trajectory analysis (TTA), the Chebyshev Rational Approximation Method (CRAM), the Quadrature-based Rational Approximation Method (QRAM) and the Laguerre Polynomial Approximation Method (LPAM). Three different modes are supported by DEPTH to execute the decay, constant flux and constant power calculations. In addition to obtaining the instantaneous quantities of the radioactivity, decay heats and reaction rates, DEPTH is able to calculate the integral quantities by a time-integrated solver. Through calculations compared with ORIGEN-2, the validity of DEPTH in point-depletion calculations is proved. The accuracy and efficiency of depletion algorithms are also discussed. In addition, an actual pin-cell burnup case is calculated to illustrate the DEPTH code performance in coupling with the RMC Monte Carlo code

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

    Bjerke, M.A.

    1983-02-01

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

  6. Summary description of the scale modular code system

    International Nuclear Information System (INIS)

    Parks, C.V.

    1987-12-01

    SCALE - a modular code system for Standardized Computer Analyses for Licensing Evaluation - has been developed at Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission staff. The SCALE system utilizes well-established computer codes and methods within standard analytic sequences that allow simplified free-form input, automate the data processing and coupling between codes, and provide accurate and reliable results. System development has been directed at criticality safety, shielding, and heat transfer analysis of spent fuel transport and/or storage casks. However, only a few of the sequences (and none of the individual functional modules) are restricted to cask applications. This report will provide a background on the history of the SCALE development and review the components and their function within the system. The available data libraries are also discussed, together with the automated features that standardize the data processing and systems analysis. 83 refs., 32 figs., 11 tabs

  7. SCALE Code System 6.2.1

    Energy Technology Data Exchange (ETDEWEB)

    Rearden, Bradley T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jessee, Matthew Anderson [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

  8. SCALE Code System 6.2.2

    Energy Technology Data Exchange (ETDEWEB)

    Rearden, Bradley T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jessee, Matthew Anderson [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.

  9. SCALE Code System 6.2.1

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Jessee, Matthew Anderson

    2016-01-01

    The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE's graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

  10. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

  11. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  12. Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

  13. Development of a new EMP code at LANL

    Science.gov (United States)

    Colman, J. J.; Roussel-Dupré, R. A.; Symbalisty, E. M.; Triplett, L. A.; Travis, B. J.

    2006-05-01

    A new code for modeling the generation of an electromagnetic pulse (EMP) by a nuclear explosion in the atmosphere is being developed. The source of the EMP is the Compton current produced by the prompt radiation (γ-rays, X-rays, and neutrons) of the detonation. As a first step in building a multi- dimensional EMP code we have written three kinetic codes, Plume, Swarm, and Rad. Plume models the transport of energetic electrons in air. The Plume code solves the relativistic Fokker-Planck equation over a specified energy range that can include ~ 3 keV to 50 MeV and computes the resulting electron distribution function at each cell in a two dimensional spatial grid. The energetic electrons are allowed to transport, scatter, and experience Coulombic drag. Swarm models the transport of lower energy electrons in air, spanning 0.005 eV to 30 keV. The swarm code performs a full 2-D solution to the Boltzmann equation for electrons in the presence of an applied electric field. Over this energy range the relevant processes to be tracked are elastic scattering, three body attachment, two body attachment, rotational excitation, vibrational excitation, electronic excitation, and ionization. All of these occur due to collisions between the electrons and neutral bodies in air. The Rad code solves the full radiation transfer equation in the energy range of 1 keV to 100 MeV. It includes effects of photo-absorption, Compton scattering, and pair-production. All of these codes employ a spherical coordinate system in momentum space and a cylindrical coordinate system in configuration space. The "z" axis of the momentum and configuration spaces is assumed to be parallel and we are currently also assuming complete spatial symmetry around the "z" axis. Benchmarking for each of these codes will be discussed as well as the way forward towards an integrated modern EMP code.

  14. Improved decoding for a concatenated coding system

    DEFF Research Database (Denmark)

    Paaske, Erik

    1990-01-01

    The concatenated coding system recommended by CCSDS (Consultative Committee for Space Data Systems) uses an outer (255,233) Reed-Solomon (RS) code based on 8-b symbols, followed by the block interleaver and an inner rate 1/2 convolutional code with memory 6. Viterbi decoding is assumed. Two new...... decoding procedures based on repeated decoding trials and exchange of information between the two decoders and the deinterleaver are proposed. In the first one, where the improvement is 0.3-0.4 dB, only the RS decoder performs repeated trials. In the second one, where the improvement is 0.5-0.6 dB, both...... decoders perform repeated decoding trials and decoding information is exchanged between them...

  15. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  16. 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)

  17. Symbol synchronization in convolutionally coded systems

    Science.gov (United States)

    Baumert, L. D.; Mceliece, R. J.; Van Tilborg, H. C. A.

    1979-01-01

    Alternate symbol inversion is sometimes applied to the output of convolutional encoders to guarantee sufficient richness of symbol transition for the receiver symbol synchronizer. A bound is given for the length of the transition-free symbol stream in such systems, and those convolutional codes are characterized in which arbitrarily long transition free runs occur.

  18. 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).

  19. 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)

  20. Coding and decoding for code division multiple user communication systems

    Science.gov (United States)

    Healy, T. J.

    1985-01-01

    A new algorithm is introduced which decodes code division multiple user communication signals. The algorithm makes use of the distinctive form or pattern of each signal to separate it from the composite signal created by the multiple users. Although the algorithm is presented in terms of frequency-hopped signals, the actual transmitter modulator can use any of the existing digital modulation techniques. The algorithm is applicable to error-free codes or to codes where controlled interference is permitted. It can be used when block synchronization is assumed, and in some cases when it is not. The paper also discusses briefly some of the codes which can be used in connection with the algorithm, and relates the algorithm to past studies which use other approaches to the same problem.

  1. Development of 1D Liner Compression Code for IDL

    Science.gov (United States)

    Shimazu, Akihisa; Slough, John; Pancotti, Anthony

    2015-11-01

    A 1D liner compression code is developed to model liner implosion dynamics in the Inductively Driven Liner Experiment (IDL) where FRC plasmoid is compressed via inductively-driven metal liners. The driver circuit, magnetic field, joule heating, and liner dynamics calculations are performed at each time step in sequence to couple these effects in the code. To obtain more realistic magnetic field results for a given drive coil geometry, 2D and 3D effects are incorporated into the 1D field calculation through use of correction factor table lookup approach. Commercial low-frequency electromagnetic fields solver, ANSYS Maxwell 3D, is used to solve the magnetic field profile for static liner condition at various liner radius in order to derive correction factors for the 1D field calculation in the code. The liner dynamics results from the code is verified to be in good agreement with the results from commercial explicit dynamics solver, ANSYS Explicit Dynamics, and previous liner experiment. The developed code is used to optimize the capacitor bank and driver coil design for better energy transfer and coupling. FRC gain calculations are also performed using the liner compression data from the code for the conceptual design of the reactor sized system for fusion energy gains.

  2. Code-modulated interferometric imaging system using phased arrays

    Science.gov (United States)

    Chauhan, Vikas; Greene, Kevin; Floyd, Brian

    2016-05-01

    Millimeter-wave (mm-wave) imaging provides compelling capabilities for security screening, navigation, and bio- medical applications. Traditional scanned or focal-plane mm-wave imagers are bulky and costly. In contrast, phased-array hardware developed for mass-market wireless communications and automotive radar promise to be extremely low cost. In this work, we present techniques which can allow low-cost phased-array receivers to be reconfigured or re-purposed as interferometric imagers, removing the need for custom hardware and thereby reducing cost. Since traditional phased arrays power combine incoming signals prior to digitization, orthogonal code-modulation is applied to each incoming signal using phase shifters within each front-end and two-bit codes. These code-modulated signals can then be combined and processed coherently through a shared hardware path. Once digitized, visibility functions can be recovered through squaring and code-demultiplexing operations. Pro- vided that codes are selected such that the product of two orthogonal codes is a third unique and orthogonal code, it is possible to demultiplex complex visibility functions directly. As such, the proposed system modulates incoming signals but demodulates desired correlations. In this work, we present the operation of the system, a validation of its operation using behavioral models of a traditional phased array, and a benchmarking of the code-modulated interferometer against traditional interferometer and focal-plane arrays.

  3. Source Code Vulnerabilities in IoT Software Systems

    Directory of Open Access Journals (Sweden)

    Saleh Mohamed Alnaeli

    2017-08-01

    Full Text Available An empirical study that examines the usage of known vulnerable statements in software systems developed in C/C++ and used for IoT is presented. The study is conducted on 18 open source systems comprised of millions of lines of code and containing thousands of files. Static analysis methods are applied to each system to determine the number of unsafe commands (e.g., strcpy, strcmp, and strlen that are well-known among research communities to cause potential risks and security concerns, thereby decreasing a system’s robustness and quality. These unsafe statements are banned by many companies (e.g., Microsoft. The use of these commands should be avoided from the start when writing code and should be removed from legacy code over time as recommended by new C/C++ language standards. Each system is analyzed and the distribution of the known unsafe commands is presented. Historical trends in the usage of the unsafe commands of 7 of the systems are presented to show how the studied systems evolved over time with respect to the vulnerable code. The results show that the most prevalent unsafe command used for most systems is memcpy, followed by strlen. These results can be used to help train software developers on secure coding practices so that they can write higher quality software systems.

  4. Bar-code automated waste tracking system

    International Nuclear Information System (INIS)

    Hull, T.E.

    1994-10-01

    The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ''stop-and-go'' operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste

  5. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Chanin, D.I.; Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems

  6. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  7. 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

  8. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  9. 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.

  10. 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

  11. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

  12. Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

  13. On the automated assessment of nuclear reactor systems code accuracy

    International Nuclear Information System (INIS)

    Kunz, Robert F.; Kasmala, Gerald F.; Mahaffy, John H.; Murray, Christopher J.

    2002-01-01

    An automated code assessment program (ACAP) has been developed to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. The tool provides a suite of metrics for quality of fit to specific data sets, and the means to produce one or more figures of merit (FOM) for a code, based on weighted averages of results from the batch execution of a large number of code-experiment and code-code data comparisons. Accordingly, this tool has the potential to significantly streamline the verification and validation (V and V) processes in NRS code development environments which are characterized by rapidly evolving software, many contributing developers and a large and growing body of validation data. In this paper, a survey of data conditioning and analysis techniques is summarized which focuses on their relevance to NRS code accuracy assessment. A number of methods are considered for their applicability to the automated assessment of the accuracy of NRS code simulations. A variety of data types and computational modeling methods are considered from a spectrum of mathematical and engineering disciplines. The goal of the survey was to identify needs, issues and techniques to be considered in the development of an automated code assessment procedure, to be used in United States Nuclear Regulatory Commission (NRC) advanced thermal-hydraulic T/H code consolidation efforts. The ACAP software was designed based in large measure on the findings of this survey. An overview of this tool is summarized and several NRS data applications are provided. The paper is organized as follows: The motivation for this work is first provided by background discussion that summarizes the relevance of this subject matter to the nuclear reactor industry. Next, the spectrum of NRS data types are classified into categories, in order to provide a basis for assessing individual comparison methods. Then, a summary of the survey is provided, where each

  14. Application of software to development of reactor-safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1980-09-01

    Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis

  15. 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.)

  16. Development of general-purpose particle and heavy ion transport monte carlo code

    International Nuclear Information System (INIS)

    Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji

    2002-01-01

    The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)

  17. Modular ORIGEN-S for multi-physics code systems

    International Nuclear Information System (INIS)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack

    2011-01-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  18. Modular ORIGEN-S for multi-physics code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2011-07-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  19. Development and assessment of the COBRA/RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  20. On the possibility of developing incoherent fibre-optic data transmission systems based on signal spectral coding with matched acousto-optical filters

    International Nuclear Information System (INIS)

    Proklov, Valerii V; Byshevski-Konopko, O A; Grigorievski, V I

    2013-01-01

    The scheme is suggested for developing the optical communication line based on the principle of code division of multiple access with matched acousto-optical filters and a 16-bit long Walsh sequence. Results of modelling show that such a line can operate if adjacent spectral lines are separated by at least double the Rayleigh criterion. (optical information transmission)

  1. 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

  2. Theoretical Atomic Physics code development II: ACE: Another collisional excitation code

    International Nuclear Information System (INIS)

    Clark, R.E.H.; Abdallah, J. Jr.; Csanak, G.; Mann, J.B.; Cowan, R.D.

    1988-12-01

    A new computer code for calculating collisional excitation data (collision strengths or cross sections) using a variety of models is described. The code uses data generated by the Cowan Atomic Structure code or CATS for the atomic structure. Collisional data are placed on a random access file and can be displayed in a variety of formats using the Theoretical Atomic Physics Code or TAPS. All of these codes are part of the Theoretical Atomic Physics code development effort at Los Alamos. 15 refs., 10 figs., 1 tab

  3. 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

  4. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  5. The PARTRAC code: Status and recent developments

    Science.gov (United States)

    Friedland, Werner; Kundrat, Pavel

    Biophysical modeling is of particular value for predictions of radiation effects due to manned space missions. PARTRAC is an established tool for Monte Carlo-based simulations of radiation track structures, damage induction in cellular DNA and its repair [1]. Dedicated modules describe interactions of ionizing particles with the traversed medium, the production and reactions of reactive species, and score DNA damage determined by overlapping track structures with multi-scale chromatin models. The DNA repair module describes the repair of DNA double-strand breaks (DSB) via the non-homologous end-joining pathway; the code explicitly simulates the spatial mobility of individual DNA ends in parallel with their processing by major repair enzymes [2]. To simulate the yields and kinetics of radiation-induced chromosome aberrations, the repair module has been extended by tracking the information on the chromosome origin of ligated fragments as well as the presence of centromeres [3]. PARTRAC calculations have been benchmarked against experimental data on various biological endpoints induced by photon and ion irradiation. The calculated DNA fragment distributions after photon and ion irradiation reproduce corresponding experimental data and their dose- and LET-dependence. However, in particular for high-LET radiation many short DNA fragments are predicted below the detection limits of the measurements, so that the experiments significantly underestimate DSB yields by high-LET radiation [4]. The DNA repair module correctly describes the LET-dependent repair kinetics after (60) Co gamma-rays and different N-ion radiation qualities [2]. First calculations on the induction of chromosome aberrations have overestimated the absolute yields of dicentrics, but correctly reproduced their relative dose-dependence and the difference between gamma- and alpha particle irradiation [3]. Recent developments of the PARTRAC code include a model of hetero- vs euchromatin structures to enable

  6. Analog system for computing sparse codes

    Science.gov (United States)

    Rozell, Christopher John; Johnson, Don Herrick; Baraniuk, Richard Gordon; Olshausen, Bruno A.; Ortman, Robert Lowell

    2010-08-24

    A parallel dynamical system for computing sparse representations of data, i.e., where the data can be fully represented in terms of a small number of non-zero code elements, and for reconstructing compressively sensed images. The system is based on the principles of thresholding and local competition that solves a family of sparse approximation problems corresponding to various sparsity metrics. The system utilizes Locally Competitive Algorithms (LCAs), nodes in a population continually compete with neighboring units using (usually one-way) lateral inhibition to calculate coefficients representing an input in an over complete dictionary.

  7. A Consistent System for Coding Laboratory Samples

    Science.gov (United States)

    Sih, John C.

    1996-07-01

    A formal laboratory coding system is presented to keep track of laboratory samples. Preliminary useful information regarding the sample (origin and history) is gained without consulting a research notebook. Since this system uses and retains the same research notebook page number for each new experiment (reaction), finding and distinguishing products (samples) of the same or different reactions becomes an easy task. Using this system multiple products generated from a single reaction can be identified and classified in a uniform fashion. Samples can be stored and filed according to stage and degree of purification, e.g. crude reaction mixtures, recrystallized samples, chromatographed or distilled products.

  8. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  9. Application bar-code system for solid radioactive waste management

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. H.; Kim, T. K.; Kang, I. S.; Cho, H. S.; Son, J. S. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    Solid radioactive wastes are generated from the post-irradiated fuel examination facility, the irradiated material examination facility, the research reactor, and the laboratories at KAERI. A bar-code system for a solid radioactive waste management of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by bar-code system.

  10. ER@CEBAF: Modeling code developments

    Energy Technology Data Exchange (ETDEWEB)

    Meot, F. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Roblin, Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)

    2016-04-13

    A proposal for a multiple-pass, high-energy, energy-recovery experiment using CEBAF is under preparation in the frame of a JLab-BNL collaboration. In view of beam dynamics investigations regarding this project, in addition to the existing model in use in Elegant a version of CEBAF is developed in the stepwise ray-tracing code Zgoubi, Beyond the ER experiment, it is also planned to use the latter for the study of polarization transport in the presence of synchrotron radiation, down to Hall D line where a 12 GeV polarized beam can be delivered. This Note briefly reports on the preliminary steps, and preliminary outcomes, based on an Elegant to Zgoubi translation.

  11. A mean field theory of coded CDMA systems

    International Nuclear Information System (INIS)

    Yano, Toru; Tanaka, Toshiyuki; Saad, David

    2008-01-01

    We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems

  12. A mean field theory of coded CDMA systems

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Toru [Graduate School of Science and Technology, Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa 223-8522 (Japan); Tanaka, Toshiyuki [Graduate School of Informatics, Kyoto University, Yoshida Hon-machi, Sakyo-ku, Kyoto-shi, Kyoto 606-8501 (Japan); Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)], E-mail: yano@thx.appi.keio.ac.jp

    2008-08-15

    We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems.

  13. ICARE/CATHARE and ASTEC code development trends

    International Nuclear Information System (INIS)

    Chatelard, P.; Dorsselaere, J.-P. van

    2000-01-01

    Regarding the computer code development for simulation of LWR severe accidents, IPSN developed a two-tier approach based on detailed codes such as ICARE/CATHARE and simplified models to be assembled in the ASTEC integral code. The ICARE/CATHARE code results from the coupling between the ICARE2 code modelling the core degradation phenomena and the thermalhydraulics code CATHARE2. It allows to calculate PWR and VVER severe accident sequences in the whole RCS. The modelling of the early degradation phase can be considered as rather complete in the ICARE/CATHARE V1 mod1 version (to be released by mid-2000) whereas some models are still missing for the late phase. The main future developments (ICARE/CATHARE V2) will concern the multi-dimensional thermalhydraulics, the quenching of partially damaged cores (mechanical and chemical effects), the debris bed two-phase thermalhydraulics (including reflooding) and the corium behaviour in the lower head. The main other physical improvements should concern the behaviour of boron carbide control rods, the processes governing the core loss of geometry (transition phase) and the oxidation of relocated melts. The ASTEC (Accident Source Term Evaluation Code) integral code, commonly developed by IPSN and GRS, aims to predict an entire LWR (PWR, VVER and BWR) severe accident sequence from the initiating event through to FP release out of the containment, for source term, PSA level 2, or accident management studies. The version ASTEC VO.3 to be released by mid-2000 can be considered now as robust and fast-running enough (between 2 and 12 hours for a one day accident) and allows to perform, with a containment multi-compartment configuration, any scenario accident study accounting for the main safety systems and operator procedures (spray, recombiner, etc.). The next version ASTEC V1, to be released beginning of 2002, will include the frontend simulation and improve modelling of in-vessel core degradation. A large validation activity will

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  15. Sequence Coding and Search System for licensee event reports: code listings. Volume 2

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2

  16. Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

    Science.gov (United States)

    Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

    2016-02-01

    The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and

  17. Progress on China nuclear data processing code system

    Science.gov (United States)

    Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

    2017-09-01

    China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

  18. 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)

  19. Simulation of water hammer phenomena using the system code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2017-07-15

    Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

  20. Simulation of water hammer phenomena using the system code ATHLET

    International Nuclear Information System (INIS)

    Bratfisch, Christoph; Koch, Marco K.

    2017-01-01

    Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

  1. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  2. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  3. Development of fast and accurate Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Mori, Takamasa

    2001-01-01

    The development work of fast and accurate Monte Carlo code MVP has started at JAERI in late 80s. From the beginning, the code was designed to utilize vector supercomputers and achieved higher computation speed by a factor of 10 or more compared with conventional codes. In 1994, the first version of MVP was released together with cross section libraries based on JENDL-3.1 and JENDL-3.2. In 1996, minor revision was made by adding several functions such as treatments of ENDF-B6 file 6 data, time dependent problem, and so on. Since 1996, several works have been carried out for the next version of MVP. The main works are (1) the development of continuous energy Monte Carlo burn-up calculation code MVP-BURN, (2) the development of a system to generate cross section libraries at arbitrary temperature, and (3) the study on error estimations and their biases in Monte Carlo eigenvalue calculations. This paper summarizes the main features of MVP, results of recent studies and future plans for MVP. (author)

  4. PHITS-a particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit

    2006-01-01

    The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range

  5. 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.

  6. 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.

  7. Development of an Auto-Validation Program for MARS Code Assessments

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2006-01-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is a best-estimate thermal hydraulic system analysis code developed at KAERI. It is important for a thermal hydraulic computer code to be assessed against theoretical and experimental data to verify and validate the performance and the integrity of the structure, models and correlations of the code. The code assessment efforts for complex thermal hydraulics code such as MARS code can be tedious, time-consuming and require large amount of human intervention in data transfer to see the results in graphic forms. Code developers produce many versions of a code during development and each version need to be verified for integrity. Thus, for MARS code developers, it is desirable to have an automatic way of carrying out the code assessment calculations. In the present work, an Auto-Validation program that carries out the code assessment efforts has been developed. The program uses the user supplied configuration file (with '.vv' extension) which contain commands to read input file, to execute the user selected MARS program, and to generate result graphs. The program can be useful if a same set of code assessments is repeated with different versions of the code. The program is written with the Delphi program language. The program runs under the Microsoft Windows environment

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  9. Library system for a one dimensional tokamak transport code: (LIBJT60), 1

    International Nuclear Information System (INIS)

    Hirayama, Toshio

    1982-12-01

    A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

  10. 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

  11. 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.)

  12. SCDAP/RELAP5/MOD3 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

    1991-01-01

    The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

  13. Development of a definitive internal dosimetry code

    International Nuclear Information System (INIS)

    Miller, G.; Inkret, W.C.; Schillaci, M.E.

    1996-01-01

    Internal dosimetry may be divided into tow main problems: (1) the forward (scientific) problem of determining biokinetics models that describe how radionuclides are taken into the body, distributed in body tissues, and excreted, and (2) the inverse (mathematical) problem: given the measured amounts in excreta and assuming a biokinetic model, to determine the times and amounts of intakes into the body. The inverse problem of internal dosimetry is, in fact, a generic problem studied in other fields (e.g., image reconstruction, spectral deconvulution, and model parameter fitting). We have developed a code for plutonium internal dosimetry using the maximum entropy method, a method for solving underdetermined inverse problems with a positivity constraint. Within the framework of Bayesian statistics, we believe the definitive approach is to examine the Bayesian posterior probability describing the probability of an intake scenario (X i ) read ( ... ) as open-quotes the set of,close quotes where X i denotes the intake amount that occurs on the with day. For plutonium, for a worker with a long employment history, this is a very high dimensional probability space, since there may be on the order of 10,000 days when intakes may have occurred. Within this high dimensional space, we calculate the mean intake scenario as i > where denotes the expectation value over the posterior probability distribution. Similarly, we calculate uncertainties and other relevant quantities, such as X 2 , as expectation values over the posterior distribution. Thanks to a recent breakthrough in describing the mathematical structure of the intake process (a Poisson sum representation of intakes), we have developed the initial version of a Bayesian expectation-value algorithm for internal dosimetry reconstructions

  14. Analysis of an XADS Target with the System Code TRACE

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor H.; Feng, Bo

    2008-01-01

    Accelerator-driven systems (ADS) present an option to reduce the radioactive waste of the nuclear industry. The experimental Accelerator-Driven System (XADS) has been designed to investigate the feasibility of using ADS on an industrial scale to burn minor actinides. The target section lies in the middle of the subcritical core and is bombarded by a proton beam to produce spallation neutrons. The thermal energy produced from this reaction requires a heat removal system for the target section. The target is cooled by liquid lead-bismuth-eutectics (LBE) in the primary system which in turn transfers the heat via a heat exchanger (HX) to the secondary coolant, Diphyl THT (DTHT), a synthetic diathermic fluid. Since this design is still in development, a detailed investigation of the system is necessary to evaluate the behavior during normal and transient operations. Due to the lack of experimental facilities and data for ADS, the analyses are mostly done using thermal hydraulic codes. In addition to evaluating the thermal hydraulics of the XADS, this paper also benchmarks a new code developed by the NRC, TRACE, against other established codes. The events used in this study are beam power switch-on/off transients and a loss of heat sink accident. The obtained results from TRACE were in good agreement with the results of various other codes. (authors)

  15. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  16. 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.

  17. 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

  18. Python-Assisted MODFLOW Application and Code Development

    Science.gov (United States)

    Langevin, C.

    2013-12-01

    The U.S. Geological Survey (USGS) has a long history of developing and maintaining free, open-source software for hydrological investigations. The MODFLOW program is one of the most popular hydrologic simulation programs released by the USGS, and it is considered to be the most widely used groundwater flow simulation code. MODFLOW was written using a modular design and a procedural FORTRAN style, which resulted in code that could be understood, modified, and enhanced by many hydrologists. The code is fast, and because it uses standard FORTRAN it can be run on most operating systems. Most MODFLOW users rely on proprietary graphical user interfaces for constructing models and viewing model results. Some recent efforts, however, have focused on construction of MODFLOW models using open-source Python scripts. Customizable Python packages, such as FloPy (https://code.google.com/p/flopy), can be used to generate input files, read simulation results, and visualize results in two and three dimensions. Automating this sequence of steps leads to models that can be reproduced directly from original data and rediscretized in space and time. Python is also being used in the development and testing of new MODFLOW functionality. New packages and numerical formulations can be quickly prototyped and tested first with Python programs before implementation in MODFLOW. This is made possible by the flexible object-oriented design capabilities available in Python, the ability to call FORTRAN code from Python, and the ease with which linear systems of equations can be solved using SciPy, for example. Once new features are added to MODFLOW, Python can then be used to automate comprehensive regression testing and ensure reliability and accuracy of new versions prior to release.

  19. Development of codes for physical calculations of WWER

    International Nuclear Information System (INIS)

    Novikov, A.N.

    2000-01-01

    A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)

  20. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    1995-01-01

    We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...

  1. Development of the coupled 'system thermal-hydraulics, 3D reactor kinetics, and hot channel' analysis capability of the MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. J.; Chung, B. D.; Lee, W.J

    2005-02-01

    The subchannel analysis capability of the MARS 3D module has been improved. Especially, the turbulent mixing and void drift models for flow mixing phenomena in rod bundles have been assessed using some well-known rod bundle test data. Then, the subchannel analysis feature was combined to the existing coupled 'system Thermal-Hydraulics (T/H) and 3D reactor kinetics' calculation capability of MARS. These features allow the coupled 'system T/H, 3D reactor kinetics, and hot channel' analysis capability and, thus, realistic simulations of hot channel behavior as well as global system T/H behavior. In this report, the MARS code features for the coupled analysis capability are described first. The code modifications relevant to the features are also given. Then, a coupled analysis of the Main Steam Line Break (MSLB) is carried out for demonstration. The results of the coupled calculations are very reasonable and realistic, and show these methods can be used to reduce the over-conservatism in the conventional safety analysis.

  2. SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

    2009-05-01

    Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)

  3. Status of development and verification of the CTFD code FLUBOX

    International Nuclear Information System (INIS)

    Graf, U.; Paradimitriou, P.

    2004-01-01

    The Computational Two-Fluid Dynamics (CTFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. FLUBOX will also be used as a multidimensional module for the German system code ATHLET. The Benchmark test cases of the European ASTAR project were used to verify the ability of the code FLUBOX to calculate typical two-phase flow phenomena and conditions: void and pressure wave propagation, phase transitions, countercurrent flows, sharp interface movements, compressible (vapour) and nearly incompressible (water) conditions, thermal and mechanical non-equilibrium, stiff source terms due to mass and heat transfer between the phases. Realistic simulations of two-phase require beside the pure conservation equations additional transport equations for the interfacial area, turbulent energy and dissipation. A transport equation for the interfacial area density covering the whole two-phase flow range is in development. First validation calculations are presented in the paper. Turbulent shear stress for two-phase flows will be modelled by the development of transport equations for the turbulent kinetic energy and the turbulent dissipation rate. The development of the transport equations is mainly based on first principles on bubbles or drops and is largely free from empiricism. (author)

  4. Development of computer code in PNC, 8

    International Nuclear Information System (INIS)

    Ohhira, Mitsuru

    1990-01-01

    Private buildings applied base isolation system, are on the practical stage now. So, under Construction and Maintenance Management Office, we are doing an application study of base isolation system to nuclear fuel facilities. On the process of this study, we have developed Dynamic Analysis Program-Base Isolation System (DAP-BS) which is able to run a 32-bit personal computer. Using this program, we can analyze a 3-dimensional structure, and evaluate the various properties of base isolation parts that are divided into maximum 16 blocks. And from the results of some simulation analyses, we thought that DAP-BS had good reliability and marketability. So, we put DAP-BS on the market. (author)

  5. Recent Developments in the Code RITRACKS (Relativistic Ion Tracks)

    Science.gov (United States)

    Plante, Ianik; Ponomarev, Artem L.; Blattnig, Steve R.

    2018-01-01

    The code RITRACKS (Relativistic Ion Tracks) was developed to simulate detailed stochastic radiation track structures of ions of different types and energies. Many new capabilities were added to the code during the recent years. Several options were added to specify the times at which the tracks appear in the irradiated volume, allowing the simulation of dose-rate effects. The code has been used to simulate energy deposition in several targets: spherical, ellipsoidal and cylindrical. More recently, density changes as well as a spherical shell were implemented for spherical targets, in order to simulate energy deposition in walled tissue equivalent proportional counters. RITRACKS is used as a part of the new program BDSTracks (Biological Damage by Stochastic Tracks) to simulate several types of chromosome aberrations in various irradiation conditions. The simulation of damage to various DNA structures (linear and chromatin fiber) by direct and indirect effects has been improved and is ongoing. Many improvements were also made to the graphic user interface (GUI), including the addition of several labels allowing changes of units. A new GUI has been added to display the electron ejection vectors. The parallel calculation capabilities, notably the pre- and post-simulation processing on Windows and Linux machines have been reviewed to make them more portable between different systems. The calculation part is currently maintained in an Atlassian Stash® repository for code tracking and possibly future collaboration.

  6. 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

  7. 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

  8. The Vulnerability Assessment Code for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Yoo, Ho Sik

    2007-01-01

    To neutralize the increasing terror threats, nuclear facilities have strong physical protection system (PPS). PPS includes detectors, door locks, fences, regular guard patrols, and a hot line to a nearest military force. To design an efficient PPS and to fully operate it, vulnerability assessment process is required. Evaluating PPS of a nuclear facility is complicate process and, hence, several assessment codes have been developed. The estimation of adversary sequence interruption (EASI) code analyzes vulnerability along a single intrusion path. To evaluate many paths to a valuable asset in an actual facility, the systematic analysis of vulnerability to intrusion (SAVI) code was developed. KAERI improved SAVI and made the Korean analysis of vulnerability to intrusion (KAVI) code. Existing codes (SAVI and KAVI) have limitations in representing the distance of a facility because they use the simplified model of a PPS called adversary sequence diagram. In adversary sequence diagram the position of doors, sensors and fences is described just as the locating area. Thus, the distance between elements is inaccurate and we cannot reflect the range effect of sensors. In this abstract, we suggest accurate and intuitive vulnerability assessment based on raster map modeling of PPS. The raster map of PPS accurately represents the relative position of elements and, thus, the range effect of sensor can be easily incorporable. Most importantly, the raster map is easy to understand

  9. Development of authentication code for multi-access optical code division multiplexing based quantum key distribution

    Science.gov (United States)

    Taiwo, Ambali; Alnassar, Ghusoon; Bakar, M. H. Abu; Khir, M. F. Abdul; Mahdi, Mohd Adzir; Mokhtar, M.

    2018-05-01

    One-weight authentication code for multi-user quantum key distribution (QKD) is proposed. The code is developed for Optical Code Division Multiplexing (OCDMA) based QKD network. A unique address assigned to individual user, coupled with degrading probability of predicting the source of the qubit transmitted in the channel offer excellent secure mechanism against any form of channel attack on OCDMA based QKD network. Flexibility in design as well as ease of modifying the number of users are equally exceptional quality presented by the code in contrast to Optical Orthogonal Code (OOC) earlier implemented for the same purpose. The code was successfully applied to eight simultaneous users at effective key rate of 32 bps over 27 km transmission distance.

  10. Scaling of Thermal-Hydraulic Phenomena and System Code Assessment

    International Nuclear Information System (INIS)

    Wolfert, K.

    2008-01-01

    In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.

  11. Development of new methods for the modeling of technical systems and result evaluation for reactor safety simulation codes. Modeling, simulation models; Entwicklung neuer Methoden zur Modellierung technischer Systeme und zur Ergebnisauswertung fuer Simulationsprogramme der Reaktorsicherheit. Modellierung, Simulationsprogramme

    Energy Technology Data Exchange (ETDEWEB)

    Cester, Francesco; Deitenbeck, Helmuth; Kuentzel, Matthias; Scheuer, Josef; Voggenberger, Thomas

    2015-04-15

    The overall objective of the project is to develop a general simulation environment for program systems used in reactor safety analysis. The simulation environment provides methods for graphical modeling and evaluation of results for the simulation models. The terms of graphical modeling and evaluation of results summarize computerized methods of pre- and postprocessing for the simulation models, which can assist the user in the execution of the simulation steps. The methods comprise CAD (''Computer Aided Design'') based input tools, interactive user interfaces for the execution of the simulation and the graphical representation and visualization of the simulation results. A particular focus was set on the requirements of the system code ATHLET. A CAD tool was developed that allows the specification of 3D geometry of the plant components and the discretization with a simulation grid. The system provides inter-faces to generate the input data of the codes and to export the data for the visualization software. The CAD system was applied for the modeling of a cooling circuit and reactor pressure vessel of a PWR. For the modeling of complex systems with many components, a general purpose graphical network editor was adapted and expanded. The editor is able to simulate networks with complex topology graphically by suitable building blocks. The network editor has been enhanced and adapted to the modeling of balance of plant and thermal fluid systems in ATHLET. For the visual display of the simulation results in the local context of the 3D geometry and the simulation grid, the open source program ParaView is applied, which is widely used for 3D visualization of field data, offering multiple options for displaying and ana-lyzing the data. New methods were developed, that allow the necessary conversion of the results of the reactor safety codes and the data of the CAD models. The trans-formed data may then be imported into ParaView and visualized. The

  12. Theoretical atomic physics code development III TAPS: A display code for atomic physics data

    International Nuclear Information System (INIS)

    Clark, R.E.H.; Abdallah, J. Jr.; Kramer, S.P.

    1988-12-01

    A large amount of theoretical atomic physics data is becoming available through use of the computer codes CATS and ACE developed at Los Alamos National Laboratory. A new code, TAPS, has been written to access this data, perform averages over terms and configurations, and display information in graphical or text form. 7 refs., 13 figs., 1 tab

  13. Coded aperture material motion detection system for the ACPR

    International Nuclear Information System (INIS)

    McArthur, D.A.; Kelly, J.G.

    1975-01-01

    Single LMFBR fuel pins are being irradiated in Sandia's Annular Core Pulsed Reactor (ACPR). In these experiments single fuel pins have been driven well into the melt and vaporization regions in transients with pulse widths of about 5 ms. The ACPR is being upgraded so that it can be used to irradiate bundles of seven LMFBR fuel pins. The coded aperture material motion detection system described is being developed for this upgraded ACPR, and has for its design goals 1 mm transverse resolution (i.e., in the axial and radial directions), depth resolution of a few cm, and time resolution of 0.1 ms. The target date for development of this system is fall 1977. The paper briefly reviews the properties of coded aperture imaging, describes one possible system for the ACPR upgrade, discusses experiments which have been performed to investigate the feasibility of such a system, and describes briefly the further work required to develop such a system. The type of coded aperture to be used has not yet been fixed, but a one-dimensional section of a Fresnel zone plate appears at this time to have significant advantages

  14. Coded aperture imaging system for nuclear fuel motion detection

    International Nuclear Information System (INIS)

    Stalker, K.T.; Kelly, J.G.

    1980-01-01

    A Coded Aperature Imaging System (CAIS) has been developed at Sandia National Laboratories to image the motion of nuclear fuel rods undergoing tests simulating accident conditions within a liquid metal fast breeder reactor. The tests require that the motion of the test fuel be monitored while it is immersed in a liquid sodium coolant precluding the use of normal optical means of imaging. However, using the fission gamma rays emitted by the fuel itself and coded aperture techniques, images with 1.5 mm radial and 5 mm axial resolution have been attained. Using an electro-optical detection system coupled to a high speed motion picture camera a time resolution of one millisecond can be achieved. This paper will discuss the application of coded aperture imaging to the problem, including the design of the one-dimensional Fresnel zone plate apertures used and the special problems arising from the reactor environment and use of high energy gamma ray photons to form the coded image. Also to be discussed will be the reconstruction techniques employed and the effect of various noise sources on system performance. Finally, some experimental results obtained using the system will be presented

  15. System code improvements for modelling passive safety systems and their validation

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

  16. SWAAM-code development and verification and application to steam generator designs

    International Nuclear Information System (INIS)

    Shin, Y.W.; Valentin, R.A.

    1990-01-01

    This paper describes the family of SWAAM codes which were developed by Argonne National Laboratory to analyze the effects of sodium-water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The paper discusses the theoretical foundations and numerical treatments on which the codes are based, followed by a description of code capabilities and limitations, verification of the codes and applications to steam generator and IHTS designs. 25 refs., 14 figs

  17. Optical code division multiple access secure communications systems with rapid reconfigurable polarization shift key user code

    Science.gov (United States)

    Gao, Kaiqiang; Wu, Chongqing; Sheng, Xinzhi; Shang, Chao; Liu, Lanlan; Wang, Jian

    2015-09-01

    An optical code division multiple access (OCDMA) secure communications system scheme with rapid reconfigurable polarization shift key (Pol-SK) bipolar user code is proposed and demonstrated. Compared to fix code OCDMA, by constantly changing the user code, the performance of anti-eavesdropping is greatly improved. The Pol-SK OCDMA experiment with a 10 Gchip/s user code and a 1.25 Gb/s user data of payload has been realized, which means this scheme has better tolerance and could be easily realized.

  18. 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)

  19. On Analyzing LDPC Codes over Multiantenna MC-CDMA System

    Directory of Open Access Journals (Sweden)

    S. Suresh Kumar

    2014-01-01

    Full Text Available Multiantenna multicarrier code-division multiple access (MC-CDMA technique has been attracting much attention for designing future broadband wireless systems. In addition, low-density parity-check (LDPC code, a promising near-optimal error correction code, is also being widely considered in next generation communication systems. In this paper, we propose a simple method to construct a regular quasicyclic low-density parity-check (QC-LDPC code to improve the transmission performance over the precoded MC-CDMA system with limited feedback. Simulation results show that the coding gain of the proposed QC-LDPC codes is larger than that of the Reed-Solomon codes, and the performance of the multiantenna MC-CDMA system can be greatly improved by these QC-LDPC codes when the data rate is high.

  20. Automatic code generation for distributed robotic systems

    International Nuclear Information System (INIS)

    Jones, J.P.

    1993-01-01

    Hetero Helix is a software environment which supports relatively large robotic system development projects. The environment supports a heterogeneous set of message-passing LAN-connected common-bus multiprocessors, but the programming model seen by software developers is a simple shared memory. The conceptual simplicity of shared memory makes it an extremely attractive programming model, especially in large projects where coordinating a large number of people can itself become a significant source of complexity. We present results from three system development efforts conducted at Oak Ridge National Laboratory over the past several years. Each of these efforts used automatic software generation to create 10 to 20 percent of the system

  1. Joint design of QC-LDPC codes for coded cooperation system with joint iterative decoding

    Science.gov (United States)

    Zhang, Shunwai; Yang, Fengfan; Tang, Lei; Ejaz, Saqib; Luo, Lin; Maharaj, B. T.

    2016-03-01

    In this paper, we investigate joint design of quasi-cyclic low-density-parity-check (QC-LDPC) codes for coded cooperation system with joint iterative decoding in the destination. First, QC-LDPC codes based on the base matrix and exponent matrix are introduced, and then we describe two types of girth-4 cycles in QC-LDPC codes employed by the source and relay. In the equivalent parity-check matrix corresponding to the jointly designed QC-LDPC codes employed by the source and relay, all girth-4 cycles including both type I and type II are cancelled. Theoretical analysis and numerical simulations show that the jointly designed QC-LDPC coded cooperation well combines cooperation gain and channel coding gain, and outperforms the coded non-cooperation under the same conditions. Furthermore, the bit error rate performance of the coded cooperation employing jointly designed QC-LDPC codes is better than those of random LDPC codes and separately designed QC-LDPC codes over AWGN channels.

  2. Progress in the development of a reactivity capability in the SAM-CE system for validating fuel management codes. Interim report

    International Nuclear Information System (INIS)

    Lichtenstein, H.; Steinberg, H.; Troubetzkoy, E.; Cohen, M.O.; Chui, C.

    1978-02-01

    The SAM-CE Monte Carlo system (for three dimensional neutron, gamma ray and electron transport) has been expanded to include a reactivity capability. The implemented code modifications have effected the following improvements: (a) Doppler broadening of ENDF/B-IV based nuclear data (including fission); (b) probability table representation for the unresolved resonance range; (c) utilization of thermal scattering law data for the moderator; (d) free gas model in the absence of thermal scattering law data; (e) generalization of the nuclear element data tape structure to facilitate data management; (f) generalization of data management routines; (g) extension of the SAM-CE Complex Combinatorial Geometry capability to facilitate treatment of hexagonal lattices; (h) simultaneous use of 4 different eigenvalue estimators; (i) estimation of the eigenfunction in user prescribed spatial domains; and (j) variance reduction via stratification of source (position, energy, direction) and absorption (based on a quota sampling technique), as well as optional suppression of absorption. The new coding has undergone extensive testing, both specific (via drivers and idealized data) and integral (via comparison with previous computations). Base data have been examined for internal consistency and checked for reasonableness. A documented TRX-1 benchmark calculation has been performed. Agreement with other calculations, as well as with experiment, has served to validate the reactivity mode of SAM-CE. Further refinement of the cross section data processing component of SAM-CE (i.e., SAM-X) is suggested

  3. Web- and system-code based, interactive, nuclear power plant simulators

    International Nuclear Information System (INIS)

    Kim, K. D.; Jain, P.; Rizwan, U.

    2006-01-01

    Using two different approaches, on-line, web- and system-code based graphical user interfaces have been developed for reactor system analysis. Both are LabVIEW (graphical programming language developed by National Instruments) based systems that allow local users as well as those at remote sites to run, interact and view the results of the system code in a web browser. In the first approach, only the data written by the system code in a tab separated ASCII output file is accessed and displayed graphically. In the second approach, LabVIEW virtual instruments are coupled with the system code as dynamic link libraries (DLL). RELAP5 is used as the system code to demonstrate the capabilities of these approaches. From collaborative projects between teams in geographically remote locations to providing system code experience to distance education students, these tools can be very beneficial in many areas of teaching and R and D. (authors)

  4. 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)

  5. Development of INCTAC code for analyzing criticality accident phenomena

    International Nuclear Information System (INIS)

    Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

    2003-01-01

    Aiming at understanding nuclear transients and thermal- and hydraulic-phenomena of the criticality accident, a code named INCTAC has been newly developed at the Institute of Nuclear Safety. The code is applicable to the analysis of criticality accident transients of aqueous homogenous fuel solution system. Neutronic transient model is composed of equations for the kinetics and for the spatial distributions, which are deduced from the time dependent multi-group transport equations with the quasi steady state assumption. Thermal-hydraulic transient model is composed of a complete set of the mass, momentum and energy equations together with the two-phase flow assumptions. Validation tests of INCTAC were made using the data obtained at TRACY, a transient experiment criticality facility of JAERI. The calculated results with INCTAC showed a very good agreement with the experiment data, except a slight discrepancy of the time when the peak of reactor power was attained. But, the discrepancy was resolved with the use of an adequate model for movement and transfer of the void in the fuel solution mostly generated by radiolysis. With a simulation model for the transport of radioactive materials through ventilation systems to the environment, INCTAC will be used as an overall safety evaluation code of the criticality accident. (author)

  6. Development status of the lattice physics code in COSINE project

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y. [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software NEKLS, North Third Ring Road, Beijing 100029 (China)

    2013-07-01

    LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

  7. Development status of the lattice physics code in COSINE project

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y.

    2013-01-01

    LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

  8. A study on the nuclear computer codes installation and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Kim, Hee Kyung; Kang, Byung Heon; Kim, Ko Ryeo; Suh, Soong Hyok; Choi, Young Gil; Lee, Jong Bok

    1990-12-01

    From 1987 a number of technical transfer related to nuclear power plant had been performed from C-E for YGN 3 and 4 construction. Among them, installation and management of the computer codes for YGN 3 and 4 fuel and nuclear steam supply system was one of the most important project. Main objectives of this project are to establish the nuclear computer code management system, to develop QA procedure for nuclear codes, to secure the nuclear code reliability and to extend techanical applicabilities including the user-oriented utility programs for nuclear codes. Contents of performing the project in this year was to produce 215 transmittal packages of nuclear codes installation including making backup magnetic tape and microfiche for software quality assurance. Lastly, for easy reference about the nuclear codes information we presented list of code names and information on the codes which were introduced from C-E. (Author)

  9. Integrated code development for studying laser driven plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takabe, Hideaki; Nagatomo, Hideo; Sunahara, Atsusi; Ohnishi, Naofumi; Naruo, Syuji; Mima, Kunioki [Osaka Univ., Suita (Japan). Inst. of Laser Engineering

    1998-03-01

    Present status and plan for developing an integrated implosion code are briefly explained by focusing on motivation, numerical scheme and issues to be developed more. Highly nonlinear stage of Rayleigh-Taylor instability of ablation front by laser irradiation has been simulated so as to be compared with model experiments. Improvement in transport and rezoning/remapping algorithms in ILESTA code is described. (author)

  10. SALT [System Analysis Language Translater]: A steady state and dynamic systems code

    International Nuclear Information System (INIS)

    Berry, G.; Geyer, H.

    1983-01-01

    SALT (System Analysis Language Translater) is a lumped parameter approach to system analysis which is totally modular. The modules are all precompiled and only the main program, which is generated by SALT, needs to be compiled for each unique system configuration. This is a departure from other lumped parameter codes where all models are written by MACROS and then compiled for each unique configuration, usually after all of the models are lumped together and sorted to eliminate undetermined variables. The SALT code contains a robust and sophisticated steady-sate finder (non-linear equation solver), optimization capability and enhanced GEAR integration scheme which makes use of sparsity and algebraic constraints. The SALT systems code has been used for various technologies. The code was originally developed for open-cycle magnetohydrodynamic (MHD) systems. It was easily extended to liquid metal MHD systems by simply adding the appropriate models and property libraries. Similarly, the model and property libraries were expanded to handle fuel cell systems, flue gas desulfurization systems, combined cycle gasification systems, fluidized bed combustion systems, ocean thermal energy conversion systems, geothermal systems, nuclear systems, and conventional coal-fired power plants. Obviously, the SALT systems code is extremely flexible to be able to handle all of these diverse systems. At present, the dynamic option has only been used for LMFBR nuclear power plants and geothermal power plants. However, it can easily be extended to other systems and can be used for analyzing control problems. 12 refs

  11. 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.

  12. 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.

  13. Review of ASME code criteria for control of primary loads on nuclear piping system branch connections and recommendations for additional development work

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Gwaltney, R.C.; Moore, S.E.

    1993-11-01

    This report collects and uses available data to reexamine the criteria for controlling primary loads in nuclear piping branch connections as expressed in Section III of the ASME Boiler and Pressure Vessel Code. In particular, the primary load stress indices given in NB-3650 and NB-3683 are reexamined. The report concludes that the present usage of the stress indices in the criteria equations should be continued. However, the complex treatment of combined branch and run moments is not supported by available information. Therefore, it is recommended that this combined loading evaluation procedure be replaced for primary loads by the separate leg evaluation procedure specified in NC/ND-3653.3(c) and NC/ND-3653.3(d). No recommendation is made for fatigue or secondary load evaluations for Class 1 piping. Further work should be done on the development of better criteria for treatment of combined branch and run moment effects

  14. GEOS Code Development Road Map - May, 2013

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Scott [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Settgast, Randolph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fu, Pengcheng [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Antoun, Tarabay [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ryerson, F. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-05-03

    GEOS is a massively parallel computational framework designed to enable HPC-based simulations of subsurface reservoir stimulation activities with the goal of optimizing current operations and evaluating innovative stimulation methods. GEOS will enable coupling of different solvers associated with the various physical processes occurring during reservoir stimulation in unique and sophisticated ways, adapted to various geologic settings, materials and stimulation methods. The overall architecture of the framework includes consistent data structures and will allow incorporation of additional physical and materials models as demanded by future applications. Along with predicting the initiation, propagation and reactivation of fractures, GEOS will also generate a seismic source term that can be linked with seismic wave propagation codes to generate synthetic microseismicity at surface and downhole arrays. Similarly, the output from GEOS can be linked with existing fluid/thermal transport codes. GEOS can also be linked with existing, non-intrusive uncertainty quantification schemes to constrain uncertainty in its predictions and sensitivity to the various parameters describing the reservoir and stimulation operations. We anticipate that an implicit-explicit 3D version of GEOS, including a preliminary seismic source model, will be available for parametric testing and validation against experimental and field data by Oct. 1, 2013.

  15. Development of covariance capabilities in EMPIRE code

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Pigni, M.T.; Oblozinsky, P.; Mughabghab, S.F.; Mattoon, C.M.; Capote, R.; Cho, Young-Sik; Trkov, A.

    2008-06-24

    The nuclear reaction code EMPIRE has been extended to provide evaluation capabilities for neutron cross section covariances in the thermal, resolved resonance, unresolved resonance and fast neutron regions. The Atlas of Neutron Resonances by Mughabghab is used as a primary source of information on uncertainties at low energies. Care is taken to ensure consistency among the resonance parameter uncertainties and those for thermal cross sections. The resulting resonance parameter covariances are formatted in the ENDF-6 File 32. In the fast neutron range our methodology is based on model calculations with the code EMPIRE combined with experimental data through several available approaches. The model-based covariances can be obtained using deterministic (Kalman) or stochastic (Monte Carlo) propagation of model parameter uncertainties. We show that these two procedures yield comparable results. The Kalman filter and/or the generalized least square fitting procedures are employed to incorporate experimental information. We compare the two approaches analyzing results for the major reaction channels on {sup 89}Y. We also discuss a long-standing issue of unreasonably low uncertainties and link it to the rigidity of the model.

  16. Computer codes and methods for simulating accelerator driven systems

    International Nuclear Information System (INIS)

    Sartori, E.; Byung Chan Na

    2003-01-01

    A large set of computer codes and associated data libraries have been developed by nuclear research and industry over the past half century. A large number of them are in the public domain and can be obtained under agreed conditions from different Information Centres. The areas covered comprise: basic nuclear data and models, reactor spectra and cell calculations, static and dynamic reactor analysis, criticality, radiation shielding, dosimetry and material damage, fuel behaviour, safety and hazard analysis, heat conduction and fluid flow in reactor systems, spent fuel and waste management (handling, transportation, and storage), economics of fuel cycles, impact on the environment of nuclear activities etc. These codes and models have been developed mostly for critical systems used for research or power generation and other technological applications. Many of them have not been designed for accelerator driven systems (ADS), but with competent use, they can be used for studying such systems or can form the basis for adapting existing methods to the specific needs of ADS's. The present paper describes the types of methods, codes and associated data available and their role in the applications. It provides Web addresses for facilitating searches for such tools. Some indications are given on the effect of non appropriate or 'blind' use of existing tools to ADS. Reference is made to available experimental data that can be used for validating the methods use. Finally, some international activities linked to the different computational aspects are described briefly. (author)

  17. Variable-length code construction for incoherent optical CDMA systems

    Science.gov (United States)

    Lin, Jen-Yung; Jhou, Jhih-Syue; Wen, Jyh-Horng

    2007-04-01

    The purpose of this study is to investigate the multirate transmission in fiber-optic code-division multiple-access (CDMA) networks. In this article, we propose a variable-length code construction for any existing optical orthogonal code to implement a multirate optical CDMA system (called as the multirate code system). For comparison, a multirate system where the lower-rate user sends each symbol twice is implemented and is called as the repeat code system. The repetition as an error-detection code in an ARQ scheme in the repeat code system is also investigated. Moreover, a parallel approach for the optical CDMA systems, which is proposed by Marić et al., is also compared with other systems proposed in this study. Theoretical analysis shows that the bit error probability of the proposed multirate code system is smaller than other systems, especially when the number of lower-rate users is large. Moreover, if there is at least one lower-rate user in the system, the multirate code system accommodates more users than other systems when the error probability of system is set below 10 -9.

  18. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  19. Development of AGNES, a kinetics code for fissile solutions, 1

    International Nuclear Information System (INIS)

    Nakajima, Ken; Ohnishi, Nobuaki

    1986-01-01

    A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

  20. A Robust Cross Coding Scheme for OFDM Systems

    NARCIS (Netherlands)

    Shao, X.; Slump, Cornelis H.

    2010-01-01

    In wireless OFDM-based systems, coding jointly over all the sub-carriers simultaneously performs better than coding separately per sub-carrier. However, the joint coding is not always optimal because its achievable channel capacity (i.e. the maximum data rate) is inversely proportional to the

  1. 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

  2. Development of tools for automatic generation of PLC code

    CERN Document Server

    Koutli, Maria; Rochez, Jacques

    This Master thesis was performed at CERN and more specifically in the EN-ICE-PLC section. The Thesis describes the integration of two PLC platforms, that are based on CODESYS development tool, to the CERN defined industrial framework, UNICOS. CODESYS is a development tool for PLC programming, based on IEC 61131-3 standard, and is adopted by many PLC manufacturers. The two PLC development environments are, the SoMachine from Schneider and the TwinCAT from Beckhoff. The two CODESYS compatible PLCs, should be controlled by the SCADA system of Siemens, WinCC OA. The framework includes a library of Function Blocks (objects) for the PLC programs and a software for automatic generation of the PLC code based on this library, called UAB. The integration aimed to give a solution that is shared by both PLC platforms and was based on the PLCOpen XML scheme. The developed tools were demonstrated by creating a control application for both PLC environments and testing of the behavior of the code of the library.

  3. Variable code gamma ray imaging system

    International Nuclear Information System (INIS)

    Macovski, A.; Rosenfeld, D.

    1979-01-01

    A gamma-ray source distribution in the body is imaged onto a detector using an array of apertures. The transmission of each aperture is modulated using a code such that the individual views of the source through each aperture can be decoded and separated. The codes are chosen to maximize the signal to noise ratio for each source distribution. These codes determine the photon collection efficiency of the aperture array. Planar arrays are used for volumetric reconstructions and circular arrays for cross-sectional reconstructions. 14 claims

  4. Channel coding in the space station data system network

    Science.gov (United States)

    Healy, T.

    1982-01-01

    A detailed discussion of the use of channel coding for error correction, privacy/secrecy, channel separation, and synchronization is presented. Channel coding, in one form or another, is an established and common element in data systems. No analysis and design of a major new system would fail to consider ways in which channel coding could make the system more effective. The presence of channel coding on TDRS, Shuttle, the Advanced Communication Technology Satellite Program system, the JSC-proposed Space Operations Center, and the proposed 30/20 GHz Satellite Communication System strongly support the requirement for the utilization of coding for the communications channel. The designers of the space station data system have to consider the use of channel coding.

  5. Recent developments in the CONTAIN-LMR code

    International Nuclear Information System (INIS)

    Murata, K.K.

    1990-01-01

    Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab

  6. 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)

  7. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  8. Alignment effects on a neutron imaging system using coded apertures

    International Nuclear Information System (INIS)

    Thfoin, Isabelle; Landoas, Olivier; Caillaud, Tony; Vincent, Maxime; Bourgade, Jean-Luc; Rosse, Bertrand; Disdier, Laurent; Sangster, Thomas C.; Glebov, Vladimir Yu.; Pien, Greg; Armstrong, William

    2010-01-01

    A high resolution neutron imaging system is being developed and tested on the OMEGA laser facility for inertial confinement fusion experiments. This diagnostic uses a coded imaging technique with a penumbral or an annular aperture. The sensitiveness of these techniques to misalignment was pointed out with both experiments and simulations. Results obtained during OMEGA shots are in good agreement with calculations performed with the Monte Carlo code GEANT4. Both techniques are sensitive to the relative position of the source in the field of view. The penumbral imaging technique then demonstrates to be less sensitive to misalignment compared to the ring. These results show the necessity to develop a neutron imaging diagnostic for megajoule class lasers taking into account our alignment capabilities on such facilities.

  9. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  10. Recent developments in KTF. Code optimization and improved numerics

    International Nuclear Information System (INIS)

    Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin

    2012-01-01

    The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

  11. Recent developments in KTF. Code optimization and improved numerics

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin [Karlsruhe Institute of Technology (KIT) (Germany). Inst. for Neutron Physics and Reactor Technology (INR)

    2012-11-01

    The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

  12. 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

  13. Twelve gordian knots when developing an organizational code of ethics

    NARCIS (Netherlands)

    Kaptein, Muel; Wempe, Johan

    1998-01-01

    Following the example of the many organizations in the United States which have a code of ethics, an increasing interest on the part of companies, trade organizations, (semi-)governmental organizations and professions in the Netherlands to develop codes of ethics can be witnessed. We have been able

  14. Developing improved MD codes for understanding processive cellulases

    International Nuclear Information System (INIS)

    Crowley, M F; Nimlos, M R; Himmel, M E; Uberbacher, E C; Iii, C L Brooks; Walker, R C

    2008-01-01

    The mechanism of action of cellulose-degrading enzymes is illuminated through a multidisciplinary collaboration that uses molecular dynamics (MD) simulations and expands the capabilities of MD codes to allow simulations of enzymes and substrates on petascale computational facilities. There is a class of glycoside hydrolase enzymes called cellulases that are thought to decrystallize and processively depolymerize cellulose using biochemical processes that are largely not understood. Understanding the mechanisms involved and improving the efficiency of this hydrolysis process through computational models and protein engineering presents a compelling grand challenge. A detailed understanding of cellulose structure, dynamics and enzyme function at the molecular level is required to direct protein engineers to the right modifications or to understand if natural thermodynamic or kinetic limits are in play. Much can be learned about processivity by conducting carefully designed molecular dynamics (MD) simulations of the binding and catalytic domains of cellulases with various substrate configurations, solvation models and thermodynamic protocols. Most of these numerical experiments, however, will require significant modification of existing code and algorithms in order to efficiently use current (terascale) and future (petascale) hardware to the degree of parallelism necessary to simulate a system of the size proposed here. This work will develop MD codes that can efficiently use terascale and petascale systems, not just for simple classical MD simulations, but also for more advanced methods, including umbrella sampling with complex restraints and reaction coordinates, transition path sampling, steered molecular dynamics, and quantum mechanical/molecular mechanical simulations of systems the size of cellulose degrading enzymes acting on cellulose

  15. The application of LDPC code in MIMO-OFDM system

    Science.gov (United States)

    Liu, Ruian; Zeng, Beibei; Chen, Tingting; Liu, Nan; Yin, Ninghao

    2018-03-01

    The combination of MIMO and OFDM technology has become one of the key technologies of the fourth generation mobile communication., which can overcome the frequency selective fading of wireless channel, increase the system capacity and improve the frequency utilization. Error correcting coding introduced into the system can further improve its performance. LDPC (low density parity check) code is a kind of error correcting code which can improve system reliability and anti-interference ability, and the decoding is simple and easy to operate. This paper mainly discusses the application of LDPC code in MIMO-OFDM system.

  16. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro

    2007-02-01

    The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

  17. RSMASS system model development

    International Nuclear Information System (INIS)

    Marshall, A.C.; Gallup, D.R.

    1998-01-01

    RSMASS system mass models have been used for more than a decade to make rapid estimates of space reactor power system masses. This paper reviews the evolution of the RSMASS models and summarizes present capabilities. RSMASS has evolved from a simple model used to make rough estimates of space reactor and shield masses to a versatile space reactor power system model. RSMASS uses unique reactor and shield models that permit rapid mass optimization calculations for a variety of space reactor power and propulsion systems. The RSMASS-D upgrade of the original model includes algorithms for the balance of the power system, a number of reactor and shield modeling improvements, and an automatic mass optimization scheme. The RSMASS-D suite of codes cover a very broad range of reactor and power conversion system options as well as propulsion and bimodal reactor systems. Reactor choices include in-core and ex-core thermionic reactors, liquid metal cooled reactors, particle bed reactors, and prismatic configuration reactors. Power conversion options include thermoelectric, thermionic, Stirling, Brayton, and Rankine approaches. Program output includes all major component masses and dimensions, efficiencies, and a description of the design parameters for a mass optimized system. In the past, RSMASS has been used as an aid to identify and select promising concepts for space power applications. The RSMASS modeling approach has been demonstrated to be a valuable tool for guiding optimization of the power system design; consequently, the model is useful during system design and development as well as during the selection process. An improved in-core thermionic reactor system model RSMASS-T is now under development. The current development of the RSMASS-T code represents the next evolutionary stage of the RSMASS models. RSMASS-T includes many modeling improvements and is planned to be more user-friendly. RSMASS-T will be released as a fully documented, certified code at the end of

  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. Quality assurance and verification of the MACCS [MELCOR Accident Consequence Code System] code, Version 1.5

    International Nuclear Information System (INIS)

    Dobbe, C.A.; Carlson, E.R.; Marshall, N.H.; Marwil, E.S.; Tolli, J.E.

    1990-02-01

    An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consequence Code System (MACCS) was performed. The QA and verification involved examination of the code and associated documentation for consistent and correct implementation of the models in an error-free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are also presented. The QA and verification process concluded that Version 1.5 of the MACCS code, within the scope and limitations process concluded that Version 1.5 of the MACCS code, within the scope and limitations of the models implemented in the code is essentially error free and ready for widespread use. 15 refs., 11 tabs

  20. Structure and operation of the ITS code system

    International Nuclear Information System (INIS)

    Halbleib, J.

    1988-01-01

    The TIGER series of time-independent coupled electron-photon Monte Carlo transport codes is a group of multimaterial and multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron-photon cascade by combining microscopic photon transport with a macroscopic random walk for electron transport. Major contributors to its evolution are listed. The author and his associates are primarily code users rather than code developers, and have borrowed freely from existing work wherever possible. Nevertheless, their efforts have resulted in various software packages for describing the production and transport of the electron-photon cascade that they found sufficiently useful to warrant dissemination through the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory. The ITS system (Integrated TIGER Series) represents the organization and integration of this combined software, along with much additional capability from previously unreleased work, into a single convenient package of exceptional user friendliness and portability. Emphasis is on simplicity and flexibility of application without sacrificing the rigor or sophistication of the physical model

  1. Development of ASME Code Section 11 visual examination requirements

    International Nuclear Information System (INIS)

    Cook, J.F.

    1990-01-01

    Section XI of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) defines three types of nondestructive examinations, visual, surface, and volumetric. Visual examination is important since it is the primary examination method for many safety-related components and systems and is also used as a backup examination for the components and systems which receive surface or volumetric examinations. Recent activity in the Section XI Code organization to improve the rules for visual examinations is reviewed and the technical basis for the new rules, which cover illumination, vision acuity, and performance demonstration, is explained

  2. Development of the code for filter calculation

    International Nuclear Information System (INIS)

    Gritzay, O.O.; Vakulenko, M.M.

    2012-01-01

    This paper describes a calculation method, which commonly used in the Neutron Physics Department to develop a new neutron filter or to improve the existing neutron filter. This calculation is the first step of the traditional filter development procedure. It allows easy selection of the qualitative and quantitative contents of a composite filter in order to receive the filtered neutron beam with given parameters

  3. Interface of RETRAN/MASTER Code System for APR1400

    International Nuclear Information System (INIS)

    Ku, Keuk Jong; Kang, Sang Hee; Kim, Han Gon

    2008-01-01

    MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is a nuclear analysis and design code which can simulate the pressurized water reactor core or boiling water reactor core in 3-dimensional geometry. RETRAN is a best-estimate code for transient analysis of Non-LOCA. RETRAN code generates neutron number density in core using point kinetics model which includes feedback reactivities and converts the neutron number density into reactor power. It is conventional that RETRAN code for power generation is roughly to extrapolate feedback reactivities which are provided by MASTER code only one time before transient analysis. The purpose of this paper is to interface RETRAN code with MASTER code by real-time processing and to supply adequate feedback reactivities to RETRAN code. So, we develop interface code called MATRAN for real-time feedback reactivity processing. And for the application of MATRAN code, we compare the results of real-time MATRAN code with those of conventional RETRAN/MASTER code

  4. Proceedings of the 8th topical meeting on nuclear code development

    International Nuclear Information System (INIS)

    1993-03-01

    The 8th Topical Meeting on Nuclear Code Development, organized by Committee on Reactor Physics and Nuclear Codes Committee of Japan Atomic Energy Research Institute (JAERI), was held at Tokai Research Establishment of JAERI, on 11th and 12th of November, 1992. In the meeting, 14 papers were presented on the topics of (1) the next generation nuclear reactor design system and (2) advances of the nuclear fuel reprocessing safety analysis codes. These papers are compiled in this proceedings. (author)

  5. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Iwasaki, Takashi; Sakai, Takaaki; Enuma, Yasuhiro; Mizuno, Tomoyasu

    2002-03-01

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

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

    Science.gov (United States)

    Sikder, Somali; Ghosh, Shila

    2018-02-01

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

  7. Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation

    Science.gov (United States)

    Pinilla, Samuel; Poveda, Juan; Arguello, Henry

    2018-03-01

    Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.

  8. Developments in the Generation and Interpretation of Wire Codes (invited paper)

    International Nuclear Information System (INIS)

    Ebi, K.L.

    1999-01-01

    Three new developments in the generation and interpretation of wire codes are discussed. First, a method was developed to computer generate wire codes using data gathered from a utility database of the local distribution system and from tax assessor records. This method was used to wire code more than 250,000 residences in the greater Denver metropolitan area. There was an approximate 75% agreement with field wire coding. Other research in Denver suggests that wire codes predict some characteristics of a residence and its neighbourhood, including age, assessed value, street layout and traffic density. A third new development is the case-specular method to study the association between wire codes and childhood cancers. Recent results from applying the method to the Savitz et al and London et al studies suggest that the associations between childhood cancer and VHCC residences were strongest for residences with a backyard rather than street service drop, and for VHCC residences with LCC speculars. (author)

  9. Noncoherent Spectral Optical CDMA System Using 1D Active Weight Two-Code Keying Codes

    Directory of Open Access Journals (Sweden)

    Bih-Chyun Yeh

    2016-01-01

    Full Text Available We propose a new family of one-dimensional (1D active weight two-code keying (TCK in spectral amplitude coding (SAC optical code division multiple access (OCDMA networks. We use encoding and decoding transfer functions to operate the 1D active weight TCK. The proposed structure includes an optical line terminal (OLT and optical network units (ONUs to produce the encoding and decoding codes of the proposed OLT and ONUs, respectively. The proposed ONU uses the modified cross-correlation to remove interferences from other simultaneous users, that is, the multiuser interference (MUI. When the phase-induced intensity noise (PIIN is the most important noise, the modified cross-correlation suppresses the PIIN. In the numerical results, we find that the bit error rate (BER for the proposed system using the 1D active weight TCK codes outperforms that for two other systems using the 1D M-Seq codes and 1D balanced incomplete block design (BIBD codes. The effective source power for the proposed system can achieve −10 dBm, which has less power than that for the other systems.

  10. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  11. 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)

  12. 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).

  13. Generalized optical code construction for enhanced and Modified Double Weight like codes without mapping for SAC-OCDMA systems

    Science.gov (United States)

    Kumawat, Soma; Ravi Kumar, M.

    2016-07-01

    Double Weight (DW) code family is one of the coding schemes proposed for Spectral Amplitude Coding-Optical Code Division Multiple Access (SAC-OCDMA) systems. Modified Double Weight (MDW) code for even weights and Enhanced Double Weight (EDW) code for odd weights are two algorithms extending the use of DW code for SAC-OCDMA systems. The above mentioned codes use mapping technique to provide codes for higher number of users. A new generalized algorithm to construct EDW and MDW like codes without mapping for any weight greater than 2 is proposed. A single code construction algorithm gives same length increment, Bit Error Rate (BER) calculation and other properties for all weights greater than 2. Algorithm first constructs a generalized basic matrix which is repeated in a different way to produce the codes for all users (different from mapping). The generalized code is analysed for BER using balanced detection and direct detection techniques.

  14. The role of the PIRT process in identifying code improvements and executing code development

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.

    1997-01-01

    In September 1988, the USNRC issued a revised ECCS rule for light water reactors that allows, as an option, the use of best estimate (BE) plus uncertainty methods in safety analysis. The key feature of this licensing option relates to quantification of the uncertainty in the determination that an NPP has a low probability of violating the safety criteria specified in 10 CFR 50. To support the 1988 licensing revision, the USNRC and its contractors developed the CSAU evaluation methodology to demonstrate the feasibility of the BE plus uncertainty approach. The PIRT process, Step 3 in the CSAU methodology, was originally formulated to support the BE plus uncertainty licensing option as executed in the CSAU approach to safety analysis. Subsequent work has shown the PIRT process to be a much more powerful tool than conceived in its original form. Through further development and application, the PIRT process has shown itself to be a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. Used early in research directed toward these objectives, PIRT results also provide the technical basis and cost effective organization for new experimental programs needed to improve the safety analysis codes for new applications. The primary purpose of this paper is to describe the generic PIRT process, including typical and common illustrations from prior applications. The secondary objective is to provide guidance to future applications of the process to help them focus, in a graded approach, on systems, components, processes and phenomena that have been common in several prior applications

  15. The role of the PIRT process in identifying code improvements and executing code development

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Boyack, B.E. [Los Alamos National Lab., NM (United States)

    1997-07-01

    In September 1988, the USNRC issued a revised ECCS rule for light water reactors that allows, as an option, the use of best estimate (BE) plus uncertainty methods in safety analysis. The key feature of this licensing option relates to quantification of the uncertainty in the determination that an NPP has a {open_quotes}low{close_quotes} probability of violating the safety criteria specified in 10 CFR 50. To support the 1988 licensing revision, the USNRC and its contractors developed the CSAU evaluation methodology to demonstrate the feasibility of the BE plus uncertainty approach. The PIRT process, Step 3 in the CSAU methodology, was originally formulated to support the BE plus uncertainty licensing option as executed in the CSAU approach to safety analysis. Subsequent work has shown the PIRT process to be a much more powerful tool than conceived in its original form. Through further development and application, the PIRT process has shown itself to be a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. Used early in research directed toward these objectives, PIRT results also provide the technical basis and cost effective organization for new experimental programs needed to improve the safety analysis codes for new applications. The primary purpose of this paper is to describe the generic PIRT process, including typical and common illustrations from prior applications. The secondary objective is to provide guidance to future applications of the process to help them focus, in a graded approach, on systems, components, processes and phenomena that have been common in several prior applications.

  16. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

  17. Development of probabilistic fracture mechanics code PASCAL and user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

    2001-03-01

    As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

  18. Development of the CRIPTE Code for Electromagnetic Coupling

    National Research Council Canada - National Science Library

    Parmantier, Jean-Philippe

    2005-01-01

    .... This code was originally developed as part of an experiment performed under the joint US-France international data exchange program on the atmospheric electricity/aircraft interactions, DEA-AF-79-7336...

  19. 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.

  20. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

    1999-01-01

    SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

  1. Ocean Thermal Energy Conversion power system development. Phase I: preliminary design. Final report. [ODSP-3 code; OTEC Steady-State Analysis Program

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-04

    The following appendices are included; Dynamic Simulation Program (ODSP-3); sample results of dynamic simulation; trip report - NH/sub 3/ safety precautions/accident records; trip report - US Coast Guard Headquarters; OTEC power system development, preliminary design test program report; medium turbine generator inspection point program; net energy analysis; bus bar cost of electricity; OTEC technical specifications; and engineer drawings. (WHK)

  2. Method of laser beam coding for control systems

    Science.gov (United States)

    Pałys, Tomasz; Arciuch, Artur; Walczak, Andrzej; Murawski, Krzysztof

    2017-08-01

    The article presents the method of encoding a laser beam for control systems. The experiments were performed using a red laser emitting source with a wavelength of λ = 650 nm and a power of P ≍ 3 mW. The aim of the study was to develop methods of modulation and demodulation of the laser beam. Results of research, in which we determined the effect of selected camera parameters, such as image resolution, number of frames per second on the result of demodulation of optical signal, is also shown in the paper. The experiments showed that the adopted coding method provides sufficient information encoded in a single laser beam (36 codes with the effectiveness of decoding at 99.9%).

  3. Mobile systems development

    DEFF Research Database (Denmark)

    Pedersen, Ole; Kristiansen, Martin Lund; Kammersgaard, Marc N.

    2007-01-01

    in XP. In general, we find XP well-suited for mobile systems development projects. However, based on our experiences and an analytical comparison we propose the following modifications to XP: Make an essential design to avoid the worst time waste during refactoring. For faster development, reuse code......Development of mobile software is Surrounded by much uncertainty. Immature software platforms on mobile clients, a highly competitive market calling for innovation, efficiency and effectiveness in the development life cycle, and lacking end-user adoption are just some of the realities facing...... development teams in the mobile software industry. By taking a process view on development of mobile systems we seek to explore the strengths and limitations of eXtreme Programming (XP) in the context of mobile software development. Following an experimental approach a mobile systems development project...

  4. Development and application of methods to characterize code uncertainty

    International Nuclear Information System (INIS)

    Wilson, G.E.; Burtt, J.D.; Case, G.S.; Einerson, J.J.; Hanson, R.G.

    1985-01-01

    The United States Nuclear Regulatory Commission sponsors both international and domestic studies to assess its safety analysis codes. The Commission staff intends to use the results of these studies to quantify the uncertainty of the codes with a statistically based analysis method. Development of the methodology is underway. The Idaho National Engineering Laboratory contributions to the early development effort, and testing of two candidate methods are the subjects of this paper

  5. Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra

    International Nuclear Information System (INIS)

    Abdallah, J. Jr.; Clark, R.E.H.

    1988-12-01

    A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab

  6. Code-first development with Entity Framework

    CERN Document Server

    Barskiy, Sergey

    2015-01-01

    This book is intended for software developers with some prior experience with the Microsoft .NET framework who want to learn how to use Entity Framework. This book will get you up and running quickly, providing many examples that illustrate all the key concepts of Entity Framework.

  7. Aeroelastic code development activities in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Wright, A.D. [National Renewable Energy Lab., Golden, Colorado (United States)

    1996-09-01

    Designing wind turbines to be fatigue resistant and to have long lifetimes at minimal cost is a major goal of the federal wind program and the wind industry in the United States. To achieve this goal, we must be able to predict critical loads for a wide variety of different wind turbines operating under extreme conditions. The codes used for wind turbine dynamic analysis must be able to analyze a wide range of different wind turbine configurations as well as rapidly predict the loads due to turbulent wind inflow with a minimal set of degrees of freedom. Code development activities in the US have taken a two-pronged approach in order to satisfy both of these criteria: (1) development of a multi-purpose code which can be used to analyze a wide variety of wind turbine configurations without having to develop new equations of motion with each configuration change, and (2) development of specialized codes with minimal sets of specific degrees of freedom for analysis of two- and three-bladed horizontal axis wind turbines and calculation of machine loads due to turbulent inflow. In the first method we have adapted a commercial multi-body dynamics simulation package for wind turbine analysis. In the second approach we are developing specialized codes with limited degrees of freedom, usually specified in the modal domain. This paper will summarize progress to date in the development, validation, and application of these codes. (au) 13 refs.

  8. 14 CFR Sec. 1-4 - System of accounts coding.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false System of accounts coding. Sec. 1-4 Section... General Accounting Provisions Sec. 1-4 System of accounts coding. (a) A four digit control number is assigned for each balance sheet and profit and loss account. Each balance sheet account is numbered...

  9. Performance Analysis of Optical Code Division Multiplex System

    Science.gov (United States)

    Kaur, Sandeep; Bhatia, Kamaljit Singh

    2013-12-01

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

  10. Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

    International Nuclear Information System (INIS)

    Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

    1979-01-01

    A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

  11. Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

    2016-11-01

    Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

  12. Report on nuclear industry quality assurance procedures for safety analysis computer code development and use

    International Nuclear Information System (INIS)

    Sheron, B.W.; Rosztoczy, Z.R.

    1980-08-01

    As a result of a request from Commissioner V. Gilinsky to investigate in detail the causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March, 1978, the staff undertook an extensive investigation of the vendor quality assurance practices applied to safety analysis computer code development and use. This investigation included inspections of code development and use practices of the four major Light Water Reactor Nuclear Steam Supply System vendors and a major reload fuel supplier. The conclusion reached by the staff as a result of the investigation is that vendor practices for code development and use are basically sound. A number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality assurance (QA) review and inspection process for computer codes and identified areas for improvement

  13. An engineering code to analyze hypersonic thermal management systems

    Science.gov (United States)

    Vangriethuysen, Valerie J.; Wallace, Clark E.

    1993-01-01

    Thermal loads on current and future aircraft are increasing and as a result are stressing the energy collection, control, and dissipation capabilities of current thermal management systems and technology. The thermal loads for hypersonic vehicles will be no exception. In fact, with their projected high heat loads and fluxes, hypersonic vehicles are a prime example of systems that will require thermal management systems (TMS) that have been optimized and integrated with the entire vehicle to the maximum extent possible during the initial design stages. This will not only be to meet operational requirements, but also to fulfill weight and performance constraints in order for the vehicle to takeoff and complete its mission successfully. To meet this challenge, the TMS can no longer be two or more entirely independent systems, nor can thermal management be an after thought in the design process, the typical pervasive approach in the past. Instead, a TMS that was integrated throughout the entire vehicle and subsequently optimized will be required. To accomplish this, a method that iteratively optimizes the TMS throughout the vehicle will not only be highly desirable, but advantageous in order to reduce the manhours normally required to conduct the necessary tradeoff studies and comparisons. A thermal management engineering computer code that is under development and being managed at Wright Laboratory, Wright-Patterson AFB, is discussed. The primary goal of the code is to aid in the development of a hypersonic vehicle TMS that has been optimized and integrated on a total vehicle basis.

  14. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

  15. Development of the CAT code - YGN 5 and 6 CVCS analysis tool

    International Nuclear Information System (INIS)

    Kim, S.W.; Sohn, S.H.; Seo, J.T.; Lee, S.K.

    1996-01-01

    The CAT code has been developed for the analysis of the Chemical and Volume Control System (CVCS) of the Yonggwang Nuclear Power Plant Units 5 and 6(YGN 5 and 6). The code is able to simulate the system behaviors in the operating conditions which should be considered in the design of the system. It has been developed as a stand alone code which can simulate CVCS in detail whenever correct system boundary conditions are provided. The code consists of two modules, i.e. control and process modules. The control module includes the models for the Pressurizer Level Control System, the Letdown Backpressure Control System, the Charging Backpressure Control System, and the Seal Injection Control System. Thermal-hydraulic responses of the system are simulated by the process module. The modeling of the system is based on a node and flowpath network. The thermal-hydraulic model is based on the assumption of homogeneous equilibrium mixture. The major system components such as valves, orifices, pumps, heat exchangers and the volume control tank are explicitly modeled in the code. The code was validated against the measured data from the letdown system test performed during the Hot Functional Testing at YGN 3. The comparison between the measured and predicted data demonstrated that the present model can predict the observed phenomena with sufficient accuracy. (author)

  16. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I

    1985-07-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.

  17. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs

  18. Development of Parallel Code for the Alaska Tsunami Forecast Model

    Science.gov (United States)

    Bahng, B.; Knight, W. R.; Whitmore, P.

    2014-12-01

    The Alaska Tsunami Forecast Model (ATFM) is a numerical model used to forecast propagation and inundation of tsunamis generated by earthquakes and other means in both the Pacific and Atlantic Oceans. At the U.S. National Tsunami Warning Center (NTWC), the model is mainly used in a pre-computed fashion. That is, results for hundreds of hypothetical events are computed before alerts, and are accessed and calibrated with observations during tsunamis to immediately produce forecasts. ATFM uses the non-linear, depth-averaged, shallow-water equations of motion with multiply nested grids in two-way communications between domains of each parent-child pair as waves get closer to coastal waters. Even with the pre-computation the task becomes non-trivial as sub-grid resolution gets finer. Currently, the finest resolution Digital Elevation Models (DEM) used by ATFM are 1/3 arc-seconds. With a serial code, large or multiple areas of very high resolution can produce run-times that are unrealistic even in a pre-computed approach. One way to increase the model performance is code parallelization used in conjunction with a multi-processor computing environment. NTWC developers have undertaken an ATFM code-parallelization effort to streamline the creation of the pre-computed database of results with the long term aim of tsunami forecasts from source to high resolution shoreline grids in real time. Parallelization will also permit timely regeneration of the forecast model database with new DEMs; and, will make possible future inclusion of new physics such as the non-hydrostatic treatment of tsunami propagation. The purpose of our presentation is to elaborate on the parallelization approach and to show the compute speed increase on various multi-processor systems.

  19. Structural reliability methods: Code development status

    Science.gov (United States)

    Millwater, Harry R.; Thacker, Ben H.; Wu, Y.-T.; Cruse, T. A.

    1991-05-01

    The Probabilistic Structures Analysis Method (PSAM) program integrates state of the art probabilistic algorithms with structural analysis methods in order to quantify the behavior of Space Shuttle Main Engine structures subject to uncertain loadings, boundary conditions, material parameters, and geometric conditions. An advanced, efficient probabilistic structural analysis software program, NESSUS (Numerical Evaluation of Stochastic Structures Under Stress) was developed as a deliverable. NESSUS contains a number of integrated software components to perform probabilistic analysis of complex structures. A nonlinear finite element module NESSUS/FEM is used to model the structure and obtain structural sensitivities. Some of the capabilities of NESSUS/FEM are shown. A Fast Probability Integration module NESSUS/FPI estimates the probability given the structural sensitivities. A driver module, PFEM, couples the FEM and FPI. NESSUS, version 5.0, addresses component reliability, resistance, and risk.

  20. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

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

    Science.gov (United States)

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

    2016-11-01

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

  2. The FLUKA code for space applications Recent developments

    CERN Document Server

    Andersen, V; Battistoni, G; Campanella, M; Carboni, M; Cerutti, F; Empl, A; Fassò, A; Ferrari, A; Gadioli, E; Garzelli, M V; Lee, K; Ottolenghi, A; Pelliccioni, M; Pinsky, L S; Ranft, J; Roesler, S; Sala, P R; Wilson, T L

    2004-01-01

    The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to- date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results h...

  3. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  4. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

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

  5. 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)

  6. PWR core follow calculations using the ELCOS code system

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1990-01-01

    The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

  7. Development of a multispectral autoradiography using a coded aperture

    Science.gov (United States)

    Noto, Daisuke; Takeda, Tohoru; Wu, Jin; Lwin, Thet T.; Yu, Quanwen; Zeniya, Tsutomu; Yuasa, Tetsuya; Hiranaka, Yukio; Itai, Yuji; Akatsuka, Takao

    2000-11-01

    Autoradiography is a useful imaging technique to understand biological functions using tracers including radio isotopes (RI's). However, it is not easy to describe the distribution of different kinds of tracers simultaneously by conventional autoradiography using X-ray film or Imaging plate. Each tracer describes each corresponding biological function. Therefore, if we can simultaneously estimate distribution of different kinds of tracer materials, the multispectral autoradiography must be a quite powerful tool to better understand physiological mechanisms of organs. So we are developing a system using a solid state detector (SSD) with high energy- resolution. Here, we introduce an imaging technique with a coded aperture to get spatial and spectral information more efficiently. In this paper, the imaging principle is described, and its validity and fundamental property are discussed by both simulation and phantom experiments with RI's such as 201Tl, 99mTc, 67Ga, and 123I.

  8. 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.

  9. 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

  10. 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)

  11. 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)

  12. The development of depletion program coupled with Monte Carlo computer code

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

    2015-01-01

    The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

  13. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  14. Further development of the computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-01

    In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

  15. Locally Minimum Storage Regenerating Codes in Distributed Cloud Storage Systems

    Institute of Scientific and Technical Information of China (English)

    Jing Wang; Wei Luo; Wei Liang; Xiangyang Liu; Xiaodai Dong

    2017-01-01

    In distributed cloud storage sys-tems, inevitably there exist multiple node fail-ures at the same time. The existing methods of regenerating codes, including minimum storage regenerating (MSR) codes and mini-mum bandwidth regenerating (MBR) codes, are mainly to repair one single or several failed nodes, unable to meet the repair need of distributed cloud storage systems. In this paper, we present locally minimum storage re-generating (LMSR) codes to recover multiple failed nodes at the same time. Specifically, the nodes in distributed cloud storage systems are divided into multiple local groups, and in each local group (4, 2) or (5, 3) MSR codes are constructed. Moreover, the grouping method of storage nodes and the repairing process of failed nodes in local groups are studied. The-oretical analysis shows that LMSR codes can achieve the same storage overhead as MSR codes. Furthermore, we verify by means of simulation that, compared with MSR codes, LMSR codes can reduce the repair bandwidth and disk I/O overhead effectively.

  16. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

  17. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  18. The Development of the World Anti-Doping Code.

    Science.gov (United States)

    Young, Richard

    2017-01-01

    This chapter addresses both the development and substance of the World Anti-Doping Code, which came into effect in 2003, as well as the subsequent Code amendments, which came into effect in 2009 and 2015. Through an extensive process of stakeholder input and collaboration, the World Anti-Doping Code has transformed the hodgepodge of inconsistent and competing pre-2003 anti-doping rules into a harmonized and effective approach to anti-doping. The Code, as amended, is now widely recognized worldwide as the gold standard in anti-doping. The World Anti-Doping Code originally went into effect on January 1, 2004. The first amendments to the Code went into effect on January 1, 2009, and the second amendments on January 1, 2015. The Code and the related international standards are the product of a long and collaborative process designed to make the fight against doping more effective through the adoption and implementation of worldwide harmonized rules and best practices. © 2017 S. Karger AG, Basel.

  19. Cooperation of experts' opinion, experiment and computer code development

    International Nuclear Information System (INIS)

    Wolfert, K.; Hicken, E.

    The connection between code development, code assessment and confidence in the analysis of transients will be discussed. In this manner, the major sources of errors in the codes and errors in applications of the codes will be shown. Standard problem results emphasize that, in order to have confidence in licensing statements, the codes must be physically realistic and the code user must be qualified and experienced. We will discuss why there is disagreement between the licensing authority and vendor concerning assessment of the fullfillment of safety goal requirements. The answer to the question lies in the different confidence levels of the assessment of transient analysis. It is expected that a decrease in the disagreement will result from an increased confidence level. Strong efforts will be made to increase this confidence level through improvements in the codes, experiments and related organizational strcutures. Because of the low probability for loss-of-coolant-accidents in the nuclear industry, assessment must rely on analytical techniques and experimental investigations. (orig./HP) [de

  20. DEVELOPMENT OF SALES APPLICATION OF PREPAID ELECTRICITY VOUCHER BASED ON ANFROID PLATFORM USING QUICK RESPONSE CODE (QR CODE

    Directory of Open Access Journals (Sweden)

    Ricky Akbar

    2017-09-01

    Full Text Available Perusahaan Listrik Negara (PLN has implemented a smart electricity system or prepaid electricity. The customers pay the electricity voucher first before use the electricity. The token contained in electricity voucher that has been purchased by the customer is inserted into the Meter Prabayar (MPB installed in the location of customers. When a customer purchases a voucher, it will get a receipt that contains all of the customer's identity and the 20-digit of voucher code (token to be entered into MPB as a substitute for electrical energy credit. Receipts obtained by the customer is certainly vulnerable to loss, or hijacked by unresponsible parties. In this study, authors designed and develop an android based application by utilizing QR code technology as a replacement for the receipt of prepaid electricity credit which contains the identity of the customer and the 20-digit voucher code. The application is developed by implemented waterfall methodology. The implementation process of the waterfall methods used, are (1 analysis of functional requirement of the system by conducting a preliminary study and data collection based on field studies and literature, (2 system design by using UML diagrams and Business Process Model Notation (BPMN and Entity Relationship diagram (ERD, (3 design implementation by using OOP (Object Oriented programming technique. Web application is developed by using laravel PHP framework and database MySQL while mobile application is developed by using B4A (4 developed system is tested by using blackbox method testing. Final result of this research is a Web and mobile applications for the sale of electricityvoucher by QR Code technology.

  1. The symbol coding language for the BUTs processor of in-core reactor control systems

    International Nuclear Information System (INIS)

    Vorob'ev, D.M.; Golovanov, M.N.; Levin, G.L.; Parfenova, T.K.; Filatov, V.P.

    1978-01-01

    A symbolic coding language is described; it has been developed for automation of making up programs for in-core control systems. The systems use the ideology of the CAMAC-VECTOR system and include the BUTs-20 processor. The symbolic coding language has been developed as a programming language of the ASSEMBLER type. Operators of instructions and pseudo-instructions, the rules of reading in the text of the source program, and operator record formats are considered

  2. Code Development for Control Design Applications: Phase I: Structural Modeling

    International Nuclear Information System (INIS)

    Bir, G. S.; Robinson, M.

    1998-01-01

    The design of integrated controls for a complex system like a wind turbine relies on a system model in an explicit format, e.g., state-space format. Current wind turbine codes focus on turbine simulation and not on system characterization, which is desired for controls design as well as applications like operating turbine model analysis, optimal design, and aeroelastic stability analysis. This paper reviews structural modeling that comprises three major steps: formation of component equations, assembly into system equations, and linearization

  3. Development of Visual CINDER Code with Visual C⧣.NET

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Oyeon [Institute for Modeling and Simulation Convergence, Daegu (Korea, Republic of)

    2016-10-15

    CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study.

  4. Development of Visual CINDER Code with Visual C⧣.NET

    International Nuclear Information System (INIS)

    Kim, Oyeon

    2016-01-01

    CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study

  5. A comparative evaluation of NDR and PSAR using the CASMO-3/MASTER code system

    International Nuclear Information System (INIS)

    Sim, Jeoung Hun; Kim, Han Gon

    2009-01-01

    In order to validate nuclear design data such as the nuclear design report (NDR) and data in preliminary (or final) safety analysis report (PSAR/FSAR) and to use data for the conceptual design of new plants, the CASMO-3/MASTER code system is selected as utility code. The nuclear design of OPR1000 and APR1400 is performed with the DIT/ROCS code system. In contrast with this design code system, the accuracy of CASMO- 3/MASTER code system has not been verified. Relatively little design data has been calculated by the CASMO-3/MASTER code system for OPR1000 and APR1400 and a bias system has not been developed yet. As such, validation of the performance of the CASMO- 3/MASTER code system is necessary. In order to validate the performance of the CASMO- 3/MASTER code system and to develop a calculation methodology, a comparative evaluation with NDR of Ulchin unit 4, cycle 1(U4C1) and the PSAR of Shinkori units 3 and 4 is carried out. The results of this evaluation are presented in this paper

  6. 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

  7. 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)

  8. 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)

  9. Concatenated coding systems employing a unit-memory convolutional code and a byte-oriented decoding algorithm

    Science.gov (United States)

    Lee, L.-N.

    1977-01-01

    Concatenated coding systems utilizing a convolutional code as the inner code and a Reed-Solomon code as the outer code are considered. In order to obtain very reliable communications over a very noisy channel with relatively modest coding complexity, it is proposed to concatenate a byte-oriented unit-memory convolutional code with an RS outer code whose symbol size is one byte. It is further proposed to utilize a real-time minimal-byte-error probability decoding algorithm, together with feedback from the outer decoder, in the decoder for the inner convolutional code. The performance of the proposed concatenated coding system is studied, and the improvement over conventional concatenated systems due to each additional feature is isolated.

  10. Development of a national code of practice for structural masonry ...

    African Journals Online (AJOL)

    The problems and constraints faced by most developing countries, particularly Ghana, in developing codes of practice for structural masonry are highlighted. The steps that must be undertaken through the coordinated efforts of the National Standards Boards, Research Institutions, Universities and Professional Bodies in the ...

  11. THIDA: code system for calculation of the exposure dose rate around a fusion device

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Igarashi, Masahito.

    1978-12-01

    A code system THIDA has been developed for calculation of the exposure dose rates around a fusion device. It consists of the following: one- and two-dimensional discrete ordinate transport codes; induced activity calculation code; activation chain, activation cross section, radionuclide gamma-ray energy/intensity and gamma-ray group constant files; and gamma ray flux to exposure dose rate conversion coefficients. (author)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  14. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  15. EquiFACS: The Equine Facial Action Coding System.

    Directory of Open Access Journals (Sweden)

    Jen Wathan

    Full Text Available Although previous studies of horses have investigated their facial expressions in specific contexts, e.g. pain, until now there has been no methodology available that documents all the possible facial movements of the horse and provides a way to record all potential facial configurations. This is essential for an objective description of horse facial expressions across a range of contexts that reflect different emotional states. Facial Action Coding Systems (FACS provide a systematic methodology of identifying and coding facial expressions on the basis of underlying facial musculature and muscle movement. FACS are anatomically based and document all possible facial movements rather than a configuration of movements associated with a particular situation. Consequently, FACS can be applied as a tool for a wide range of research questions. We developed FACS for the domestic horse (Equus caballus through anatomical investigation of the underlying musculature and subsequent analysis of naturally occurring behaviour captured on high quality video. Discrete facial movements were identified and described in terms of the underlying muscle contractions, in correspondence with previous FACS systems. The reliability of others to be able to learn this system (EquiFACS and consistently code behavioural sequences was high--and this included people with no previous experience of horses. A wide range of facial movements were identified, including many that are also seen in primates and other domestic animals (dogs and cats. EquiFACS provides a method that can now be used to document the facial movements associated with different social contexts and thus to address questions relevant to understanding social cognition and comparative psychology, as well as informing current veterinary and animal welfare practices.

  16. 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

  17. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  18. Variable weight Khazani-Syed code using hybrid fixed-dynamic technique for optical code division multiple access system

    Science.gov (United States)

    Anas, Siti Barirah Ahmad; Seyedzadeh, Saleh; Mokhtar, Makhfudzah; Sahbudin, Ratna Kalos Zakiah

    2016-10-01

    Future Internet consists of a wide spectrum of applications with different bit rates and quality of service (QoS) requirements. Prioritizing the services is essential to ensure that the delivery of information is at its best. Existing technologies have demonstrated how service differentiation techniques can be implemented in optical networks using data link and network layer operations. However, a physical layer approach can further improve system performance at a prescribed received signal quality by applying control at the bit level. This paper proposes a coding algorithm to support optical domain service differentiation using spectral amplitude coding techniques within an optical code division multiple access (OCDMA) scenario. A particular user or service has a varying weight applied to obtain the desired signal quality. The properties of the new code are compared with other OCDMA codes proposed for service differentiation. In addition, a mathematical model is developed for performance evaluation of the proposed code using two different detection techniques, namely direct decoding and complementary subtraction.

  19. Geometrical Approach to the Grid System in the KOPEC Pilot Code

    International Nuclear Information System (INIS)

    Lee, E. J.; Park, C. E.; Lee, S. Y.

    2008-01-01

    KOPEC has been developing a pilot code to analyze two phase flow. The earlier version of the pilot code adopts the geometry with one-dimensional structured mesh system. As the pilot code is required to handle more complex geometries, a systematic geometrical approach to grid system has been introduced. Grid system can be classified as two types; structured grid system and unstructured grid system. The structured grid system is simple to apply but is less flexible than the other. The unstructured grid system is more complicated than the structured grid system. But it is more flexible to model the geometry. Therefore, two types of grid systems are utilized to allow code users simplicity as well as the flexibility

  20. On transform coding tools under development for VP10

    Science.gov (United States)

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

    2016-09-01

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

  1. Optimization of multi-phase compressible lattice Boltzmann codes on massively parallel multi-core systems

    NARCIS (Netherlands)

    Biferale, L.; Mantovani, F.; Pivanti, M.; Pozzati, F.; Sbragaglia, M.; Schifano, S.F.; Toschi, F.; Tripiccione, R.

    2011-01-01

    We develop a Lattice Boltzmann code for computational fluid-dynamics and optimize it for massively parallel systems based on multi-core processors. Our code describes 2D multi-phase compressible flows. We analyze the performance bottlenecks that we find as we gradually expose a larger fraction of

  2. Applications of the lots computer code to laser fusion systems and other physical optics problems

    International Nuclear Information System (INIS)

    Lawrence, G.; Wolfe, P.N.

    1979-01-01

    The Laser Optical Train Simulation (LOTS) code has been developed at the Optical Sciences Center, University of Arizona under contract to Los Alamos Scientific Laboratory (LASL). LOTS is a diffraction based code designed to beam quality and energy of the laser fusion system in an end-to-end calculation