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

Integrated computer codes for nuclear power plant severe accident analysis

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Integrated computer codes for nuclear power plant severe accident analysis

International Nuclear Information System (INIS)

Jordanov, I.; Khristov, Y.

1995-01-01

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

Verification study of the FORE-2M nuclear/thermal-hydraulilc analysis computer code

International Nuclear Information System (INIS)

Coffield, R.D.; Tang, Y.S.; Markley, R.A.

1982-01-01

The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments. (orig.)

Qualification of FEAST 3.0 and FEAT 4.0 computer codes

International Nuclear Information System (INIS)

Xu, Z.; Lai, L.; Sim, K.-S.; Huang, F.; Wong, B.

2005-01-01

FEAST (Finite Element Analysis for Stresses) is an AECL computer code used to assess the structural integrity of the CANDU fuel element. FEAST models the thermo-elastic, thermo-elasto-plastic and creep deformations in CANDU fuel. FEAT (Finite Element Analysis for Temperature) is another AECL computer code and is used to assess the thermal integrity of fuel elements. FEAT models the steady-state and transient heat flows in CANDU fuel, under conditions such as flux depression, end flux peaking, temperature-dependent thermal conductivity, and non-uniform time-dependent boundary conditions. Both computer programs are used in design and qualification analyses of CANDU fuel. Formal qualifications (including coding verification and validation) of both codes were performed, in accordance with AECL software quality assurance (SQA) manual and procedures that are consistent with CSA N286.7-99. Validation of FEAST 3.0 shows very good agreement with independent analytical solutions or measurements. Validation of FEAT 4.0 also shows very good agreement with independent WIMS-AECL calculations, analytical solutions, ANSYS calculations and measurement. (author)

Qualification of FEAST 3.0 and FEAT 4.0 computer codes

Energy Technology Data Exchange (ETDEWEB)

Xu, Z.; Lai, L.; Sim, K.-S.; Huang, F.; Wong, B. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

2005-07-01

FEAST (Finite Element Analysis for Stresses) is an AECL computer code used to assess the structural integrity of the CANDU fuel element. FEAST models the thermo-elastic, thermo-elasto-plastic and creep deformations in CANDU fuel. FEAT (Finite Element Analysis for Temperature) is another AECL computer code and is used to assess the thermal integrity of fuel elements. FEAT models the steady-state and transient heat flows in CANDU fuel, under conditions such as flux depression, end flux peaking, temperature-dependent thermal conductivity, and non-uniform time-dependent boundary conditions. Both computer programs are used in design and qualification analyses of CANDU fuel. Formal qualifications (including coding verification and validation) of both codes were performed, in accordance with AECL software quality assurance (SQA) manual and procedures that are consistent with CSA N286.7-99. Validation of FEAST 3.0 shows very good agreement with independent analytical solutions or measurements. Validation of FEAT 4.0 also shows very good agreement with independent WIMS-AECL calculations, analytical solutions, ANSYS calculations and measurement. (author)

SKYSHIN: A computer code for calculating radiation dose over a barrier

International Nuclear Information System (INIS)

Atwood, C.L.; Boland, J.R.; Dickman, P.T.

1986-11-01

SKYSHIN is a computer code for calculating the radioactive dose (mrem), when there is a barrier between the point source and the receptor. The two geometrical configurations considered are: the source and receptor separated by a rectangular wall, and the source at the bottom of a cylindrical hole in the ground. Each gamma ray traveling over the barrier is assumed to be scattered at a single point. The dose to a receptor from such paths is numerically integrated for the total dose, with symmetry used to reduce the triple integral to a double integral. The buildup factor used along a straight line through air is based on published data, and extrapolated in a stable way to low energy levels. This buildup factor was validated by comparing calculated and experimental line-of-sight doses. The entire code shows good agreement to limited field data. The code runs on a CDC or on a Vax computer, and could be modified easily for others

Computer code validation by high temperature chemistry

International Nuclear Information System (INIS)

Alexander, C.A.; Ogden, J.S.

1988-01-01

At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered

Development of an integral computer code for simulation of heat exchangers

International Nuclear Information System (INIS)

Horvat, A.; Catton, I.

2001-01-01

Heat exchangers are one of the basic installations in power and process industries. The present guidelines provide an ad-hoc solution to certain design problems. A unified approach based on simultaneous modeling of thermal-hydraulics and structural behavior does not exist. The present paper describes the development of integral numerical code for simulation of heat exchangers. The code is based on Volume Averaging Technique (VAT) for porous media flow modeling. The calculated values of the whole-section drag and heat transfer coefficients show an excellent agreement with already published values. The matching results prove the correctness of the selected approach and verify the developed numerical code used for this calculation.(author)

European Validation of the Integral Code ASTEC (EVITA)

International Nuclear Information System (INIS)

Allelein, H.-J.; Neu, K.; Dorsselaere, J.P. Van

2005-01-01

The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC (Accident Source Term Evaluation Code)-commonly developed and basically validated by GRS and IRSN-was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:- for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases; - for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users' tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme

Basic data, computer codes and integral experiments: The tools for modelling in nuclear technology

International Nuclear Information System (INIS)

Sartori, E.

2001-01-01

When studying applications in nuclear technology we need to understand and be able to predict the behavior of systems manufactured by human enterprise. First, the underlying basic physical and chemical phenomena need to be understood. We have then to predict the results from the interplay of the large number of the different basic events: i.e. the macroscopic effects. In order to be able to build confidence in our modelling capability, we need then to compare these results against measurements carried out on such systems. The different levels of modelling require the solution of different types of equations using different type of parameters. The tools required for carrying out a complete validated analysis are: - The basic nuclear or chemical data; - The computer codes, and; - The integral experiments. This article describes the role each component plays in a computational scheme designed for modelling purposes. It describes also which tools have been developed and are internationally available. The role of the OECD/NEA Data Bank, the Radiation Shielding Information Computational Center (RSICC), and the IAEA Nuclear Data Section are playing in making these elements available to the community of scientists and engineers is described. (author)

Quantum computation with Turaev-Viro codes

International Nuclear Information System (INIS)

Koenig, Robert; Kuperberg, Greg; Reichardt, Ben W.

2010-01-01

For a 3-manifold with triangulated boundary, the Turaev-Viro topological invariant can be interpreted as a quantum error-correcting code. The code has local stabilizers, identified by Levin and Wen, on a qudit lattice. Kitaev's toric code arises as a special case. The toric code corresponds to an abelian anyon model, and therefore requires out-of-code operations to obtain universal quantum computation. In contrast, for many categories, such as the Fibonacci category, the Turaev-Viro code realizes a non-abelian anyon model. A universal set of fault-tolerant operations can be implemented by deforming the code with local gates, in order to implement anyon braiding. We identify the anyons in the code space, and present schemes for initialization, computation and measurement. This provides a family of constructions for fault-tolerant quantum computation that are closely related to topological quantum computation, but for which the fault tolerance is implemented in software rather than coming from a physical medium.

Computing Challenges in Coded Mask Imaging

Science.gov (United States)

Skinner, Gerald

2009-01-01

This slide presaentation reviews the complications and challenges in developing computer systems for Coded Mask Imaging telescopes. The coded mask technique is used when there is no other way to create the telescope, (i.e., when there are wide fields of view, high energies for focusing or low energies for the Compton/Tracker Techniques and very good angular resolution.) The coded mask telescope is described, and the mask is reviewed. The coded Masks for the INTErnational Gamma-Ray Astrophysics Laboratory (INTEGRAL) instruments are shown, and a chart showing the types of position sensitive detectors used for the coded mask telescopes is also reviewed. Slides describe the mechanism of recovering an image from the masked pattern. The correlation with the mask pattern is described. The Matrix approach is reviewed, and other approaches to image reconstruction are described. Included in the presentation is a review of the Energetic X-ray Imaging Survey Telescope (EXIST) / High Energy Telescope (HET), with information about the mission, the operation of the telescope, comparison of the EXIST/HET with the SWIFT/BAT and details of the design of the EXIST/HET.

Geochemical computer codes. A review

International Nuclear Information System (INIS)

Andersson, K.

1987-01-01

In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)

Survey of computer codes applicable to waste facility performance evaluations

International Nuclear Information System (INIS)

Alsharif, M.; Pung, D.L.; Rivera, A.L.; Dole, L.R.

1988-01-01

This study is an effort to review existing information that is useful to develop an integrated model for predicting the performance of a radioactive waste facility. A summary description of 162 computer codes is given. The identified computer programs address the performance of waste packages, waste transport and equilibrium geochemistry, hydrological processes in unsaturated and saturated zones, and general waste facility performance assessment. Some programs also deal with thermal analysis, structural analysis, and special purposes. A number of these computer programs are being used by the US Department of Energy, the US Nuclear Regulatory Commission, and their contractors to analyze various aspects of waste package performance. Fifty-five of these codes were identified as being potentially useful on the analysis of low-level radioactive waste facilities located above the water table. The code summaries include authors, identification data, model types, and pertinent references. 14 refs., 5 tabs

CBP Phase I Code Integration

International Nuclear Information System (INIS)

Smith, F.; Brown, K.; Flach, G.; Sarkar, S.

2011-01-01

The goal of the Cementitious Barriers Partnership (CBP) is to develop a reasonable and credible set of software tools to predict the structural, hydraulic, and chemical performance of cement barriers used in nuclear applications over extended time frames (greater than 100 years for operating facilities and greater than 1000 years for waste management). The simulation tools will be used to evaluate and predict the behavior of cementitious barriers used in near surface engineered waste disposal systems including waste forms, containment structures, entombments, and environmental remediation. These cementitious materials are exposed to dynamic environmental conditions that cause changes in material properties via (i) aging, (ii) chloride attack, (iii) sulfate attack, (iv) carbonation, (v) oxidation, and (vi) primary constituent leaching. A set of state-of-the-art software tools has been selected as a starting point to capture these important aging and degradation phenomena. Integration of existing software developed by the CBP partner organizations was determined to be the quickest method of meeting the CBP goal of providing a computational tool that improves the prediction of the long-term behavior of cementitious materials. These partner codes were selected based on their maturity and ability to address the problems outlined above. The GoldSim Monte Carlo simulation program (GTG 2010a, GTG 2010b) was chosen as the code integration platform (Brown and Flach 2009b). GoldSim (current Version 10.5) is a Windows based graphical object-oriented computer program that provides a flexible environment for model development (Brown and Flach 2009b). The linking of GoldSim to external codes has previously been successfully demonstrated (Eary 2007, Mattie et al. 2007). GoldSim is capable of performing deterministic and probabilistic simulations and of modeling radioactive decay and constituent transport. As part of the CBP project, a general Dynamic Link Library (DLL) interface

CBP PHASE I CODE INTEGRATION

Energy Technology Data Exchange (ETDEWEB)

Smith, F.; Brown, K.; Flach, G.; Sarkar, S.

2011-09-30

The goal of the Cementitious Barriers Partnership (CBP) is to develop a reasonable and credible set of software tools to predict the structural, hydraulic, and chemical performance of cement barriers used in nuclear applications over extended time frames (greater than 100 years for operating facilities and greater than 1000 years for waste management). The simulation tools will be used to evaluate and predict the behavior of cementitious barriers used in near surface engineered waste disposal systems including waste forms, containment structures, entombments, and environmental remediation. These cementitious materials are exposed to dynamic environmental conditions that cause changes in material properties via (i) aging, (ii) chloride attack, (iii) sulfate attack, (iv) carbonation, (v) oxidation, and (vi) primary constituent leaching. A set of state-of-the-art software tools has been selected as a starting point to capture these important aging and degradation phenomena. Integration of existing software developed by the CBP partner organizations was determined to be the quickest method of meeting the CBP goal of providing a computational tool that improves the prediction of the long-term behavior of cementitious materials. These partner codes were selected based on their maturity and ability to address the problems outlined above. The GoldSim Monte Carlo simulation program (GTG 2010a, GTG 2010b) was chosen as the code integration platform (Brown & Flach 2009b). GoldSim (current Version 10.5) is a Windows based graphical object-oriented computer program that provides a flexible environment for model development (Brown & Flach 2009b). The linking of GoldSim to external codes has previously been successfully demonstrated (Eary 2007, Mattie et al. 2007). GoldSim is capable of performing deterministic and probabilistic simulations and of modeling radioactive decay and constituent transport. As part of the CBP project, a general Dynamic Link Library (DLL) interface was

TRAC-CFD code integration and its application to containment analysis

International Nuclear Information System (INIS)

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

2004-01-01

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

Algorithms and computer codes for atomic and molecular quantum scattering theory

International Nuclear Information System (INIS)

Thomas, L.

1979-01-01

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

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

Science.gov (United States)

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

1997-02-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code

International Nuclear Information System (INIS)

Glaeser, H.

1995-01-01

The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)

A restructuring of CF package for MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

CF package, which evaluates user-specified 'control functions' and applies them to define or control various aspects of computation, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the CF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory, difficulty is more over because its data is location information of other package's data due to characteristics of CF package. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

Theory of the space-dependent fuel management computer code ''UAFCC''

International Nuclear Information System (INIS)

El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

1981-01-01

This report displays the theory of the spatial burnup computer code ''UAFCC'' which has been constructed as a part of an integrated reactor calculation scheme proposed at the Reactors Department of the ARE Atomic Energy Authority. The ''UAFCC'' is a single energy-one-dimensional diffusion burnup FORTRAN computer code for well moderated, multiregion, cylindrical thermal reactors. The effect of reactivity variation with burnup is introduced in the steady state diffusion equation by a fictitious neutron source. The infinite multiplication factor, the total migration area, and the power density per unit thermal flux are calculated from the point model burnup code ''UABUC'' fitted to polynomials of suitable degree in the flux-time, and then used as an input data to the ''UAFCC'' code. The proposed burnup spatial model has been used to study the different stratogemes of the incore fuel management schemes. The conclusions of this study will be presented in a future publication. (author)

Computer codes in particle transport physics

International Nuclear Information System (INIS)

Pesic, M.

2004-01-01

Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option

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

Computer codes for evaluation of control room habitability (HABIT)

International Nuclear Information System (INIS)

Stage, S.A.

1996-06-01

This report describes the Computer Codes for Evaluation of Control Room Habitability (HABIT). HABIT is a package of computer codes designed to be used for the evaluation of control room habitability in the event of an accidental release of toxic chemicals or radioactive materials. Given information about the design of a nuclear power plant, a scenario for the release of toxic chemicals or radionuclides, and information about the air flows and protection systems of the control room, HABIT can be used to estimate the chemical exposure or radiological dose to control room personnel. HABIT is an integrated package of several programs that previously needed to be run separately and required considerable user intervention. This report discusses the theoretical basis and physical assumptions made by each of the modules in HABIT and gives detailed information about the data entry windows. Sample runs are given for each of the modules. A brief section of programming notes is included. A set of computer disks will accompany this report if the report is ordered from the Energy Science and Technology Software Center. The disks contain the files needed to run HABIT on a personal computer running DOS. Source codes for the various HABIT routines are on the disks. Also included are input and output files for three demonstration runs

A 1ST Step Integration of the Restructured MELCOR for the MIDAS Computer Code

International Nuclear Information System (INIS)

Park, S. H.; Kim, D. H.; Cho, S. W.

2006-01-01

KAERI is developing a localized severe accident code, MIDAS, based on MELCOR. MELCOR uses pointer variables for a fixed-size storage management to save the data. It passes data through two depths, its meaning is not understandable by variable itself. So it is needed to understand the methods for data passing. This method deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring for each package was developed and tested. And then integration of each restructured package was being processed one by one. In this paper, the integrating scope includes the BUR, CF, CVH, DCH, EDF, ESF, MP, SPR, TF and TP packages. As most of them use data within each package and a few packages share data with other packages. The verification was done through comparing the results before and after the restructuring

Recent progress of an integrated implosion code and modeling of element physics

International Nuclear Information System (INIS)

Nagatomo, H.; Takabe, H.; Mima, K.; Ohnishi, N.; Sunahara, A.; Takeda, T.; Nishihara, K.; Nishiguchu, A.; Sawada, K.

2001-01-01

Physics of the inertial fusion is based on a variety of elements such as compressible hydrodynamics, radiation transport, non-ideal equation of state, non-LTE atomic process, and relativistic laser plasma interaction. In addition, implosion process is not in stationary state and fluid dynamics, energy transport and instabilities should be solved simultaneously. In order to study such complex physics, an integrated implosion code including all physics important in the implosion process should be developed. The details of physics elements should be studied and the resultant numerical modeling should be installed in the integrated code so that the implosion can be simulated with available computer within realistic CPU time. Therefore, this task can be basically separated into two parts. One is to integrate all physics elements into a code, which is strongly related to the development of hydrodynamic equation solver. We have developed 2-D integrated implosion code which solves mass, momentum, electron energy, ion energy, equation of states, laser ray-trace, laser absorption radiation, surface tracing and so on. The reasonable results in simulating Rayleigh-Taylor instability and cylindrical implosion are obtained using this code. The other is code development on each element physics and verification of these codes. We had progress in developing a nonlocal electron transport code and 2 and 3 dimension radiation hydrodynamic code. (author)

Translation of ARAC computer codes

International Nuclear Information System (INIS)

Takahashi, Kunio; Chino, Masamichi; Honma, Toshimitsu; Ishikawa, Hirohiko; Kai, Michiaki; Imai, Kazuhiko; Asai, Kiyoshi

1982-05-01

In 1981 we have translated the famous MATHEW, ADPIC and their auxiliary computer codes for CDC 7600 computer version to FACOM M-200's. The codes consist of a part of the Atmospheric Release Advisory Capability (ARAC) system of Lawrence Livermore National Laboratory (LLNL). The MATHEW is a code for three-dimensional wind field analysis. Using observed data, it calculates the mass-consistent wind field of grid cells by a variational method. The ADPIC is a code for three-dimensional concentration prediction of gases and particulates released to the atmosphere. It calculates concentrations in grid cells by the particle-in-cell method. They are written in LLLTRAN, i.e., LLNL Fortran language and are implemented on the CDC 7600 computers of LLNL. In this report, i) the computational methods of the MATHEW/ADPIC and their auxiliary codes, ii) comparisons of the calculated results with our JAERI particle-in-cell, and gaussian plume models, iii) translation procedures from the CDC version to FACOM M-200's, are described. Under the permission of LLNL G-Division, this report is published to keep the track of the translation procedures and to serve our JAERI researchers for comparisons and references of their works. (author)

Integral transport computation of gamma detector response with the CPM2 code

International Nuclear Information System (INIS)

Jones, D.B.

1989-12-01

CPM-2 Version 3 is an enhanced version of the CPM-2 lattice physics computer code which supports the capabilities to (1) perform a two-dimensional gamma flux calculation and (2) perform Restart/Data file maintenance operations. The Gamma Calculation Module implemented in CPM-2 was first developed for EPRI in the CASMO-1 computer code by Studsvik Energiteknik under EPRI Agreement RP2352-01. The gamma transport calculation uses the CPM-HET code module to calculate the transport of gamma rays in two dimensions in a mixed cylindrical-rectangular geometry, where the basic fuel assembly and component regions are maintained in a rectangular geometry, but the fuel pins are represented as cylinders within a square pin cell mesh. Such a capability is needed to represent gamma transport in an essentially transparent medium containing spatially distributed ''black'' cylindrical pins. Under a subcontract to RP2352-01, RPI developed the gamma production and gamma interaction library used for gamma calculation. The CPM-2 gamma calculation was verified against reference results generated by Studsvik using the CASMO-1 program. The CPM-2 Restart/Data file maintenance capabilities provide the user with options to copy files between Restart/Data tapes and to purge files from the Restart/Data tapes

Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes

International Nuclear Information System (INIS)

Chang, Y.W.; Gvildys, J.

1977-01-01

In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs

Double folding model of nucleus-nucleus potential: formulae, iteration method and computer code

International Nuclear Information System (INIS)

Luk'yanov, K.V.

2008-01-01

Method of construction of the nucleus-nucleus double folding potential is described. Iteration procedure for the corresponding integral equation is presented. Computer code and numerical results are presented

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.

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.

Evaluation of the FRAPCON-3 Computer Code

Energy Technology Data Exchange (ETDEWEB)

Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala (Sweden)

2002-03-01

, computational methods and code structure, and (ii) an integral assessment report, which presents the performed code calibration with experimental data. In addition, there are two supporting documents related to the MATPRO materials properties package, which is extensively used in FRAPCON-3. Our evaluation confirms the applicability of FRAPCON-3 to high burnup fuel rods, but also reveals weaknesses in several models. In particular, the models for UO{sub 2} thermal conductivity degradation with increasing burnup, clad oxidation, creep, plasticity and irradiation-induced axial growth have been found to be less adequate. Suggestions for improvements to these models are pointed out in the report. In conclusion, we believe that FRAPCON-3 constitutes a suitable computer code for steady-state fuel rod analysis, into which SKI can add new and improved models that satisfy their requirements.

Evaluation of the FRAPCON-3 Computer Code

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali

2002-03-01

, computational methods and code structure, and (ii) an integral assessment report, which presents the performed code calibration with experimental data. In addition, there are two supporting documents related to the MATPRO materials properties package, which is extensively used in FRAPCON-3. Our evaluation confirms the applicability of FRAPCON-3 to high burnup fuel rods, but also reveals weaknesses in several models. In particular, the models for UO 2 thermal conductivity degradation with increasing burnup, clad oxidation, creep, plasticity and irradiation-induced axial growth have been found to be less adequate. Suggestions for improvements to these models are pointed out in the report. In conclusion, we believe that FRAPCON-3 constitutes a suitable computer code for steady-state fuel rod analysis, into which SKI can add new and improved models that satisfy their requirements

Porting plasma physics simulation codes to modern computing architectures using the libmrc framework

Science.gov (United States)

Germaschewski, Kai; Abbott, Stephen

2015-11-01

Available computing power has continued to grow exponentially even after single-core performance satured in the last decade. The increase has since been driven by more parallelism, both using more cores and having more parallelism in each core, e.g. in GPUs and Intel Xeon Phi. Adapting existing plasma physics codes is challenging, in particular as there is no single programming model that covers current and future architectures. We will introduce the open-source libmrc framework that has been used to modularize and port three plasma physics codes: The extended MHD code MRCv3 with implicit time integration and curvilinear grids; the OpenGGCM global magnetosphere model; and the particle-in-cell code PSC. libmrc consolidates basic functionality needed for simulations based on structured grids (I/O, load balancing, time integrators), and also introduces a parallel object model that makes it possible to maintain multiple implementations of computational kernels, on e.g. conventional processors and GPUs. It handles data layout conversions and enables us to port performance-critical parts of a code to a new architecture step-by-step, while the rest of the code can remain unchanged. We will show examples of the performance gains and some physics applications.

Adaptation of HAMMER computer code to CYBER 170/750 computer

International Nuclear Information System (INIS)

Pinheiro, A.M.B.S.; Nair, R.P.K.

1982-01-01

The adaptation of HAMMER computer code to CYBER 170/750 computer is presented. The HAMMER code calculates cell parameters by multigroup transport theory and reactor parameters by few group diffusion theory. The auxiliary programs, the carried out modifications and the use of HAMMER system adapted to CYBER 170/750 computer are described. (M.C.K.) [pt

Integrated Fuel-Coolant Interaction (IFCI 6.0) code

International Nuclear Information System (INIS)

Davis, F.J.; Young, M.F.

1994-04-01

The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

Benchmarking of computer codes and approaches for modeling exposure scenarios

International Nuclear Information System (INIS)

Seitz, R.R.; Rittmann, P.D.; Wood, M.I.; Cook, J.R.

1994-08-01

The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

2005-07-01

Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

Neutron shielding point kernel integral calculation code for personal computer: PKN-pc

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.

1994-07-01

A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)

Verification of SACI-2 computer code comparing with experimental results of BIBLIS-A and LOOP-7 computer code

International Nuclear Information System (INIS)

Soares, P.A.; Sirimarco, L.F.

1984-01-01

SACI-2 is a computer code created to study the dynamic behaviour of a PWR nuclear power plant. To evaluate the quality of its results, SACI-2 was used to recalculate commissioning tests done in BIBLIS-A nuclear power plant and to calculate postulated transients for Angra-2 reactor. The results of SACI-2 computer code from BIBLIS-A showed as much good agreement as those calculated with the KWU Loop 7 computer code for Angra-2. (E.G.) [pt

APC: A new code for Atmospheric Polarization Computations

International Nuclear Information System (INIS)

Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.

2013-01-01

A new polarized radiative transfer code Atmospheric Polarization Computations (APC) is described. The code is based on separation of the diffuse light field into anisotropic and smooth (regular) parts. The anisotropic part is computed analytically. The smooth regular part is computed numerically using the discrete ordinates method. Vertical stratification of the atmosphere, common types of bidirectional surface reflection and scattering by spherical particles or spheroids are included. A particular consideration is given to computation of the bidirectional polarization distribution function (BPDF) of the waved ocean surface. -- Highlights: •A new code, APC, has been developed. •The code was validated against well-known codes. •The BPDF for an arbitrary Mueller matrix is computed

The RETRAN-03 computer code

International Nuclear Information System (INIS)

Paulsen, M.P.; McFadden, J.H.; Peterson, C.E.; McClure, J.A.; Gose, G.C.; Jensen, P.J.

1991-01-01

The RETRAN-03 code development effort is designed to overcome the major theoretical and practical limitations associated with the RETRAN-02 computer code. The major objectives of the development program are to extend the range of analyses that can be performed with RETRAN, to make the code more dependable and faster running, and to have a more transportable code. The first two objectives are accomplished by developing new models and adding other models to the RETRAN-02 base code. The major model additions for RETRAN-03 are as follows: implicit solution methods for the steady-state and transient forms of the field equations; additional options for the velocity difference equation; a new steady-state initialization option for computer low-power steam generator initial conditions; models for nonequilibrium thermodynamic conditions; and several special-purpose models. The source code and the environmental library for RETRAN-03 are written in standard FORTRAN 77, which allows the last objective to be fulfilled. Some models in RETRAN-02 have been deleted in RETRAN-03. In this paper the changes between RETRAN-02 and RETRAN-03 are reviewed

Computation of the bounce-average code

International Nuclear Information System (INIS)

Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.

1977-01-01

The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended

Implatation of MC2 computer code

International Nuclear Information System (INIS)

Seehusen, J.; Nair, R.P.K.; Becceneri, J.C.

1981-01-01

The implantation of MC2 computer code in the CDC system is presented. The MC2 computer code calculates multigroup cross sections for tipical compositions of fast reactors. The multigroup constants are calculated using solutions of PI or BI approximations for determined buckling value as weighting function. (M.C.K.) [pt

Cloud Computing for Complex Performance Codes.

Energy Technology Data Exchange (ETDEWEB)

Appel, Gordon John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klein, Brandon Thorin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miner, John Gifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-02-01

This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.

Computer codes for safety analysis

International Nuclear Information System (INIS)

Holland, D.F.

1986-11-01

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

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)

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).

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.

Development of integrated computer code for analysis of risk reduction strategy

International Nuclear Information System (INIS)

Kim, Dong Ha; Kim, See Darl; Kim, Hee Dong

2002-05-01

The development of the MIDAS/TH integrated severe accident code was performed in three main areas: 1) addition of new models derived from the national experimental programs and models for APR-1400 Korea next generation reactor, 2) improvement of the existing models using the recently available results, and 3) code restructuring for user friendliness. The unique MIDAS/TH models include: 1) a kinetics module for core power calculation during ATWS, 2) a gap cooling module between the molten corium pool and the reactor vessel wall, 3) a penetration tube failure module, 4) a PAR analysis module, and 5) a look-up table for the pressure and dynamic load during steam explosion. The improved models include: 1) a debris dispersal module considering the cavity geometry during DCH, 2) hydrogen burn and deflagration-to-detonation transition criteria, 3) a peak pressure estimation module for hydrogen detonation, and 4) the heat transfer module between the molten corium pool and the overlying water. The sparger and the ex-vessel heat transfer module were assessed. To enhance user friendliness, code restructuring was performed. In addition, a sample of severe accident analysis results was organized under the preliminary database structure

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)

Implementation of computer codes for performance assessment of the Republic repository of LLW/ILW Mochovce

International Nuclear Information System (INIS)

Hanusik, V.; Kopcani, I.; Gedeon, M.

2000-01-01

This paper describes selection and adaptation of computer codes required to assess the effects of radionuclide release from Mochovce Radioactive Waste Disposal Facility. The paper also demonstrates how these codes can be integrated into performance assessment methodology. The considered codes include DUST-MS for source term release, MODFLOW for ground-water flow and BS for transport through biosphere and dose assessment. (author)

Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

International Nuclear Information System (INIS)

Ito, Masahiro; Uwaba, Tomoyuki

2005-04-01

JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

Computer codes for RF cavity design

International Nuclear Information System (INIS)

Ko, K.

1992-08-01

In RF cavity design, numerical modeling is assuming an increasingly important role with the help of sophisticated computer codes and powerful yet affordable computers. A description of the cavity codes in use in the accelerator community has been given previously. The present paper will address the latest developments and discuss their applications to cavity toning and matching problems

Computer codes for RF cavity design

International Nuclear Information System (INIS)

Ko, K.

1992-01-01

In RF cavity design, numerical modeling is assuming an increasingly important role with the help of sophisticated computer codes and powerful yet affordable computers. A description of the cavity codes in use in the accelerator community has been given previously. The present paper will address the latest developments and discuss their applications to cavity tuning and matching problems. (Author) 8 refs., 10 figs

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.

Microgravity computing codes. User's guide

Science.gov (United States)

1982-01-01

Codes used in microgravity experiments to compute fluid parameters and to obtain data graphically are introduced. The computer programs are stored on two diskettes, compatible with the floppy disk drives of the Apple 2. Two versions of both disks are available (DOS-2 and DOS-3). The codes are written in BASIC and are structured as interactive programs. Interaction takes place through the keyboard of any Apple 2-48K standard system with single floppy disk drive. The programs are protected against wrong commands given by the operator. The programs are described step by step in the same order as the instructions displayed on the monitor. Most of these instructions are shown, with samples of computation and of graphics.

Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

CONCEPT computer code

International Nuclear Information System (INIS)

Delene, J.

1984-01-01

CONCEPT is a computer code that will provide conceptual capital investment cost estimates for nuclear and coal-fired power plants. The code can develop an estimate for construction at any point in time. Any unit size within the range of about 400 to 1300 MW electric may be selected. Any of 23 reference site locations across the United States and Canada may be selected. PWR, BWR, and coal-fired plants burning high-sulfur and low-sulfur coal can be estimated. Multiple-unit plants can be estimated. Costs due to escalation/inflation and interest during construction are calculated

40 CFR 194.23 - Models and computer codes.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e., computer...

A restructuring of COR package for MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

The COR package, which calculates the thermal response of the core and the lower plenum internal structures and models the relocation of the core and lower plenum structural materials, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the COR package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as a waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the COR package addressed in this paper includes a module development, subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerated the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

Development of fast ignition integrated interconnecting code (FI3) for fast ignition scheme

International Nuclear Information System (INIS)

Nagatomo, H.; Johzaki, T.; Mima, K.; Sunahara, A.; Nishihara, K.; Izawa, Y.; Sakagami, H.; Nakao, Y.; Yokota, T.; Taguchi, T.

2005-01-01

The numerical simulation plays an important role in estimating the feasibility and performance of the fast ignition. There are two key issues in numerical analysis for the fast ignition. One is the controlling the implosion dynamics to form a high density core plasma in non-spherical implosion, and the other is heating core plasma efficiency by the short pulse high intense laser. From initial laser irradiation to final fusion burning, all the physics are coupling strongly in any phase, and they must be solved consistently in computational simulation. However, in general, it is impossible to simulate laser plasma interaction and radiation hydrodynamics in a single computational code, without any numerical dissipation, special assumption or conditional treatment. Recently, we have developed 'Fast Ignition Integrated Interconnecting code' (FI 3 ) which consists of collective Particle-in-Cell code, Relativistic Fokker-Planck hydro code, and 2-dimensional radiation hydrodynamics code. And those codes are connecting with each other in data-flow bases. In this paper, we will present detail feature of the FI 3 code, and numerical results of whole process of fast ignition. (author)

Parameters that affect parallel processing for computational electromagnetic simulation codes on high performance computing clusters

Science.gov (United States)

Moon, Hongsik

What is the impact of multicore and associated advanced technologies on computational software for science? Most researchers and students have multicore laptops or desktops for their research and they need computing power to run computational software packages. Computing power was initially derived from Central Processing Unit (CPU) clock speed. That changed when increases in clock speed became constrained by power requirements. Chip manufacturers turned to multicore CPU architectures and associated technological advancements to create the CPUs for the future. Most software applications benefited by the increased computing power the same way that increases in clock speed helped applications run faster. However, for Computational ElectroMagnetics (CEM) software developers, this change was not an obvious benefit - it appeared to be a detriment. Developers were challenged to find a way to correctly utilize the advancements in hardware so that their codes could benefit. The solution was parallelization and this dissertation details the investigation to address these challenges. Prior to multicore CPUs, advanced computer technologies were compared with the performance using benchmark software and the metric was FLoting-point Operations Per Seconds (FLOPS) which indicates system performance for scientific applications that make heavy use of floating-point calculations. Is FLOPS an effective metric for parallelized CEM simulation tools on new multicore system? Parallel CEM software needs to be benchmarked not only by FLOPS but also by the performance of other parameters related to type and utilization of the hardware, such as CPU, Random Access Memory (RAM), hard disk, network, etc. The codes need to be optimized for more than just FLOPs and new parameters must be included in benchmarking. In this dissertation, the parallel CEM software named High Order Basis Based Integral Equation Solver (HOBBIES) is introduced. This code was developed to address the needs of the

A research on the verification of models used in the computational codes and the uncertainty reduction method for the containment integrity evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Moo Hwan; Seo, Kyoung Woo [POSTECH, Pohang (Korea, Republic of)

2001-03-15

In the probability approach, the calculated CCFPs of all the scenarios were zero, which meant that it was expected that for all the accident scenarios the maximum pressure load induced by DCH was lower than the containment failure pressure obtained from the fragility curve. Thus, it can be stated that the KSNP containment is robust to the DCH threat. And uncertainty of computer codes used to be two (deterministic and probabilistic) approaches were reduced by the sensitivity tests and the research with the verification and comparison of the DCH models in each code. So, this research was to evaluate synthetic result of DCH issue and expose accurate methodology to assess containment integrity about operating PWR in Korea.

Just-in-Time Compilation-Inspired Methodology for Parallelization of Compute Intensive Java Code

Directory of Open Access Journals (Sweden)

GHULAM MUSTAFA

2017-01-01

Full Text Available Compute intensive programs generally consume significant fraction of execution time in a small amount of repetitive code. Such repetitive code is commonly known as hotspot code. We observed that compute intensive hotspots often possess exploitable loop level parallelism. A JIT (Just-in-Time compiler profiles a running program to identify its hotspots. Hotspots are then translated into native code, for efficient execution. Using similar approach, we propose a methodology to identify hotspots and exploit their parallelization potential on multicore systems. Proposed methodology selects and parallelizes each DOALL loop that is either contained in a hotspot method or calls a hotspot method. The methodology could be integrated in front-end of a JIT compiler to parallelize sequential code, just before native translation. However, compilation to native code is out of scope of this work. As a case study, we analyze eighteen JGF (Java Grande Forum benchmarks to determine parallelization potential of hotspots. Eight benchmarks demonstrate a speedup of up to 7.6x on an 8-core system

Just-in-time compilation-inspired methodology for parallelization of compute intensive java code

International Nuclear Information System (INIS)

Mustafa, G.; Ghani, M.U.

2017-01-01

Compute intensive programs generally consume significant fraction of execution time in a small amount of repetitive code. Such repetitive code is commonly known as hotspot code. We observed that compute intensive hotspots often possess exploitable loop level parallelism. A JIT (Just-in-Time) compiler profiles a running program to identify its hotspots. Hotspots are then translated into native code, for efficient execution. Using similar approach, we propose a methodology to identify hotspots and exploit their parallelization potential on multicore systems. Proposed methodology selects and parallelizes each DOALL loop that is either contained in a hotspot method or calls a hotspot method. The methodology could be integrated in front-end of a JIT compiler to parallelize sequential code, just before native translation. However, compilation to native code is out of scope of this work. As a case study, we analyze eighteen JGF (Java Grande Forum) benchmarks to determine parallelization potential of hotspots. Eight benchmarks demonstrate a speedup of up to 7.6x on an 8-core system. (author)

Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

International Nuclear Information System (INIS)

Young, Michael F.

1999-01-01

The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks

Computer code MLCOSP for multiple-correlation and spectrum analysis with a hybrid computer

International Nuclear Information System (INIS)

Oguma, Ritsuo; Fujii, Yoshio; Usui, Hozumi; Watanabe, Koichi

1975-10-01

Usage of the computer code MLCOSP(Multiple Correlation and Spectrum) developed is described for a hybrid computer installed in JAERI Functions of the hybrid computer and its terminal devices are utilized ingeniously in the code to reduce complexity of the data handling which occurrs in analysis of the multivariable experimental data and to perform the analysis in perspective. Features of the code are as follows; Experimental data can be fed to the digital computer through the analog part of the hybrid computer by connecting with a data recorder. The computed results are displayed in figures, and hardcopies are taken when necessary. Series-messages to the code are shown on the terminal, so man-machine communication is possible. And further the data can be put in through a keyboard, so case study according to the results of analysis is possible. (auth.)

Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

Science.gov (United States)

Piro, Markus Hans Alexander

Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system

Development of a 3-D flow analysis computer program for integral reactor

International Nuclear Information System (INIS)

Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

2003-01-01

A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

Turbo Pascal Computer Code for PIXE Analysis

International Nuclear Information System (INIS)

Darsono

2002-01-01

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

Users guide for NRC145-2 accident assessment computer code

International Nuclear Information System (INIS)

Pendergast, M.M.

1982-08-01

An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output

Status of the ASTEC integral code

International Nuclear Information System (INIS)

Van Dorsselaere, J.P.; Jacq, F.; Allelein, H.J.

2000-01-01

The ASTEC (Accident Source Term Evaluation Code) integrated code is developed since 1997 in close collaboration by IPSN and GRS to predict an entire LWR severe accident sequence from the initiating event up to Fission Product (FP) release out of the containment. The applications of such a code are source term determination studies, scenario evaluations, accident management studies and Probabilistic Safety Assessment level 2 (PSA-2) studies. The version V0 of ASTEC is based on the RCS modules of the ESCADRE integrated code (IPSN) and on the upgraded RALOC and FIPLOC codes (GRS) for containment thermalhydraulics and aerosol behaviour. The latest version V0.2 includes the general feed-back from the overall validation performed in 1998 (25 separate-effect experiments, PHEBUS.FP FPT1 integrated experiment), some modelling improvements (i.e. silver-iodine reactions in the containment sump), and the implementation of the main safety systems for Severe Accident Management. Several reactor-applications are under way on French and German PWR, and on VVER-1000, all with a multi-compartment configuration of the containment. The total IPSN-GRS manpower involved in ASTEC project is today about 20 men/year. The main evolution of the next version V1, foreseen end of 2001, concerns the integration of the front-end phase and the improvement of the in-vessel degradation late-phase modelling. (author)

Evaluation of the SCANAIR Computer Code

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali

2001-11-01

The SCANAIR computer code, version 3.2, has been evaluated from the standpoint of its capability to analyze, simulate and predict nuclear fuel behavior during severe power transients. SCANAIR calculates the thermal and mechanical behavior of a pressurized water reactor (PWR) fuel rod during a postulated reactivity initiated accident (RIA), and our evaluation indicates that SCANAIR is a state of the art computational tool for this purpose. Our evaluation starts by reviewing the basic theoretical models in SCANAIR, namely the governing equations for heat transfer, the mechanical response of fuel and clad, and the fission gas release behavior. The numerical methods used to solve the governing equations are briefly reviewed, and the range of applicability of the models and their limitations are discussed and illustrated with examples. Next, the main features of the SCANAIR user interface are delineated. The code requires an extensive amount of input data, in order to define burnup-dependent initial conditions to the simulated RIA. These data must be provided in a special format by a thermal-mechanical fuel rod analysis code. The user also has to supply the transient power history under RIA as input, which requires a code for neutronics calculation. The programming structure and documentation of the code are also addressed in our evaluation. SCANAIR is programmed in Fortran-77, and makes use of several general Fortran-77 libraries for handling input/output, data storage and graphical presentation of computed results. The documentation of SCANAIR and its helping libraries is generally of good quality. A drawback with SCANAIR in its present form, is that the code and its pre- and post-processors are tied to computers running the Unix or Linux operating systems. As part of our evaluation, we have performed a large number of computations with SCANAIR, some of which are documented in this report. The computations presented here include a hypothetical RIA in a high

Computer codes used in particle accelerator design: First edition

International Nuclear Information System (INIS)

1987-01-01

This paper contains a listing of more than 150 programs that have been used in the design and analysis of accelerators. Given on each citation are person to contact, classification of the computer code, publications describing the code, computer and language runned on, and a short description of the code. Codes are indexed by subject, person to contact, and code acronym

MISER-I: a computer code for JOYO fuel management

International Nuclear Information System (INIS)

Yamashita, Yoshioki

1976-06-01

A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

Hamor-2: a computer code for LWR inventory calculation

International Nuclear Information System (INIS)

Guimaraes, L.N.F.; Marzo, M.A.S.

1985-01-01

A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

Modeling of the WWER-1000 fuel-rod behavior in steady-state condition with FRAPCONE-3 computer code

International Nuclear Information System (INIS)

Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

2008-01-01

It is presented within the paper the results of the modeling and the assessment of the integral code predictions of the WWER fuel-rod behavior in steady-state condition. The assessments in this paper have used the MASSIH and ANS 5.4 subroutine in the code. The modeling and calculations have been performed with FRAPCONE-3 computer code in Argonne National Laboratory, USA

A conceptual design of multidisciplinary-integrated C.F.D. simulation on parallel computers

International Nuclear Information System (INIS)

Onishi, Ryoichi; Ohta, Takashi; Kimura, Toshiya.

1996-11-01

A design of a parallel aeroelastic code for aircraft integrated simulations is conducted. The method for integrating aerodynamics and structural dynamics software on parallel computers is devised by using the Euler/Navier-Stokes equations coupled with wing-box finite element structures. A synthesis of modern aircraft requires the optimizations of aerodynamics, structures, controls, operabilities, or other design disciplines, and the R and D efforts to implement Multidisciplinary Design Optimization environments using high performance computers are made especially among the U.S. aerospace industries. This report describes a Multiple Program Multiple Data (MPMD) parallelization of aerodynamics and structural dynamics codes with a dynamic deformation grid. A three-dimensional computation of a flowfield with dynamic deformation caused by a structural deformation is performed, and a pressure data calculated is used for a computation of the structural deformation which is input again to a fluid dynamics code. This process is repeated exchanging the computed data of pressures and deformations between flowfield grids and structural elements. It enables to simulate the structure movements which take into account of the interaction of fluid and structure. The conceptual design for achieving the aforementioned various functions is reported. Also the future extensions to incorporate control systems, which enable to simulate a realistic aircraft configuration to be a major tool for Aircraft Integrated Simulation, are investigated. (author)

Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

Energy Technology Data Exchange (ETDEWEB)

Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

1985-04-01

This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

Computer codes for level 1 probabilistic safety assessment

International Nuclear Information System (INIS)

1990-06-01

Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs

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)

The Light-Water-Reactor Version of the URANUS Integral fuel-rod code

Energy Technology Data Exchange (ETDEWEB)

Labmann, K; Moreno, A

1977-07-01

The LWR version of the URANUS code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via URANUS results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analyses of in-pile experiments, they illustrate rather typical and diverse URANUS capabilities. The performance test shows that URANUS is reliable and efficient, thus the code is a most valuable tool in fuel rod analysis work. K. LaBmann developed the LWR version of the URANUS code, material properties were reviewed and supplied by A. Moreno. (Author) 41 refs.

The light-water-reactor version of the Uranus integral fuel-rod code

International Nuclear Information System (INIS)

Moreno, A.; Lassmann, K.

1977-01-01

The LWR of the Uranus code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via Uranus results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analysis of in-pile experiments, they illustrate rather typical and diverse Uranus capabilities. The performance test shows that Uranus is reliable and efficient, thus the code is a most valuable tool in fuel fod analysis work. K. Lassmann developed the LWR version of the Uranus code, material properties were reviewed and supplied by A. Moreno. (author)

On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

International Nuclear Information System (INIS)

El-Osery, I.A.

1981-01-01

The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented

HUDU: The Hanford Unified Dose Utility computer code

International Nuclear Information System (INIS)

Scherpelz, R.I.

1991-02-01

The Hanford Unified Dose Utility (HUDU) computer program was developed to provide rapid initial assessment of radiological emergency situations. The HUDU code uses a straight-line Gaussian atmospheric dispersion model to estimate the transport of radionuclides released from an accident site. For dose points on the plume centerline, it calculates internal doses due to inhalation and external doses due to exposure to the plume. The program incorporates a number of features unique to the Hanford Site (operated by the US Department of Energy), including a library of source terms derived from various facilities' safety analysis reports. The HUDU code was designed to run on an IBM-PC or compatible personal computer. The user interface was designed for fast and easy operation with minimal user training. The theoretical basis and mathematical models used in the HUDU computer code are described, as are the computer code itself and the data libraries used. Detailed instructions for operating the code are also included. Appendices to the report contain descriptions of the program modules, listings of HUDU's data library, and descriptions of the verification tests that were run as part of the code development. 14 refs., 19 figs., 2 tabs

Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

International Nuclear Information System (INIS)

Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

2010-01-01

Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

A restructuring of RN1 package for MIDAS computer code

International Nuclear Information System (INIS)

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

2003-01-01

RN1 package, which is one of two fission product-related packages in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN1 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

A restructuring of RN2 package for MIDAS computer code

International Nuclear Information System (INIS)

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

2003-01-01

RN2 package, which is one of two fission product-related package in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN2 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN2 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The validation has been done by comparing the results of the modified code with those from the existing code. As the trends are the similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

Nonuniform code concatenation for universal fault-tolerant quantum computing

Science.gov (United States)

Nikahd, Eesa; Sedighi, Mehdi; Saheb Zamani, Morteza

2017-09-01

Using transversal gates is a straightforward and efficient technique for fault-tolerant quantum computing. Since transversal gates alone cannot be computationally universal, they must be combined with other approaches such as magic state distillation, code switching, or code concatenation to achieve universality. In this paper we propose an alternative approach for universal fault-tolerant quantum computing, mainly based on the code concatenation approach proposed in [T. Jochym-O'Connor and R. Laflamme, Phys. Rev. Lett. 112, 010505 (2014), 10.1103/PhysRevLett.112.010505], but in a nonuniform fashion. The proposed approach is described based on nonuniform concatenation of the 7-qubit Steane code with the 15-qubit Reed-Muller code, as well as the 5-qubit code with the 15-qubit Reed-Muller code, which lead to two 49-qubit and 47-qubit codes, respectively. These codes can correct any arbitrary single physical error with the ability to perform a universal set of fault-tolerant gates, without using magic state distillation.

Computer codes for ventilation in nuclear facilities

International Nuclear Information System (INIS)

Mulcey, P.

1987-01-01

In this paper the authors present some computer codes, developed in the last years, for ventilation and radioprotection. These codes are used for safety analysis in the conception, exploitation and dismantlement of nuclear facilities. The authors present particularly: DACC1 code used for aerosol deposit in sampling circuit of radiation monitors; PIAF code used for modelization of complex ventilation system; CLIMAT 6 code used for optimization of air conditioning system [fr

Automated uncertainty analysis methods in the FRAP computer codes

International Nuclear Information System (INIS)

Peck, S.O.

1980-01-01

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

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)

Implantation of FRAPCON-2 code in HB computer

International Nuclear Information System (INIS)

Silva, C.F. da.

1987-05-01

The modifications carried out for implanting FRAPCON-2 computer code in the HB DPS-T7 computer are presented. The FRAPCON-2 code calculates thermo-mechanical response during long period of burnup in stationary state for fuel rods of PWR type reactors. (M.C.K.)

Overview of Grid Codes for Photovoltaic Integration

DEFF Research Database (Denmark)

Zheng, Qianwei; Li, Jiaming; Ai, Xiaomeng

2017-01-01

The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration in differ......The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration...

Computer code FIT

International Nuclear Information System (INIS)

Rohmann, D.; Koehler, T.

1987-02-01

This is a description of the computer code FIT, written in FORTRAN-77 for a PDP 11/34. FIT is an interactive program to decude position, width and intensity of lines of X-ray spectra (max. length of 4K channels). The lines (max. 30 lines per fit) may have Gauss- or Voigt-profile, as well as exponential tails. Spectrum and fit can be displayed on a Tektronix terminal. (orig.) [de

Overview of Grid Codes for Photovoltaic Integration

DEFF Research Database (Denmark)

Zheng, Qianwei; Li, Jiaming; Ai, Xiaomeng

2017-01-01

The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration in differ...... in different grid codes are first investigated. On this basis, the future advocacy is concluded. Finally, several evaluation indices are proposed to quantify the grid code compliance so that the system operators can validate all these requirements by simulation....

Non-linear heat transfer computer code by finite element method

International Nuclear Information System (INIS)

Nagato, Kotaro; Takikawa, Noboru

1977-01-01

The computer code THETA-2D for the calculation of temperature distribution by the two-dimensional finite element method was made for the analysis of heat transfer in a high temperature structure. Numerical experiment was performed for the numerical integration of the differential equation of heat conduction. The Runge-Kutta method of the numerical experiment produced an unstable solution. A stable solution was obtained by the β method with the β value of 0.35. In high temperature structures, the radiative heat transfer can not be neglected. To introduce a term of the radiative heat transfer, a functional neglecting the radiative heat transfer was derived at first. Then, the radiative term was added after the discretion by variation method. Five model calculations were carried out by the computer code. Calculation of steady heat conduction was performed. When estimated initial temperature is 1,000 degree C, reasonable heat blance was obtained. In case of steady-unsteady temperature calculation, the time integral by THETA-2D turned out to be under-estimation for enthalpy change. With a one-dimensional model, the temperature distribution in a structure, in which heat conductivity is dependent on temperature, was calculated. Calculation with a model which has a void inside was performed. Finally, model calculation for a complex system was carried out. (Kato, T.)

APC: A New Code for Atmospheric Polarization Computations

Science.gov (United States)

Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.

2014-01-01

A new polarized radiative transfer code Atmospheric Polarization Computations (APC) is described. The code is based on separation of the diffuse light field into anisotropic and smooth (regular) parts. The anisotropic part is computed analytically. The smooth regular part is computed numerically using the discrete ordinates method. Vertical stratification of the atmosphere, common types of bidirectional surface reflection and scattering by spherical particles or spheroids are included. A particular consideration is given to computation of the bidirectional polarization distribution function (BPDF) of the waved ocean surface.

Quantum computing with Majorana fermion codes

Science.gov (United States)

Litinski, Daniel; von Oppen, Felix

2018-05-01

We establish a unified framework for Majorana-based fault-tolerant quantum computation with Majorana surface codes and Majorana color codes. All logical Clifford gates are implemented with zero-time overhead. This is done by introducing a protocol for Pauli product measurements with tetrons and hexons which only requires local 4-Majorana parity measurements. An analogous protocol is used in the fault-tolerant setting, where tetrons and hexons are replaced by Majorana surface code patches, and parity measurements are replaced by lattice surgery, still only requiring local few-Majorana parity measurements. To this end, we discuss twist defects in Majorana fermion surface codes and adapt the technique of twist-based lattice surgery to fermionic codes. Moreover, we propose a family of codes that we refer to as Majorana color codes, which are obtained by concatenating Majorana surface codes with small Majorana fermion codes. Majorana surface and color codes can be used to decrease the space overhead and stabilizer weight compared to their bosonic counterparts.

TERRA: a computer code for simulating the transport of environmentally released radionuclides through agriculture

International Nuclear Information System (INIS)

Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.

1984-11-01

TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location

TERRA: a computer code for simulating the transport of environmentally released radionuclides through agriculture

Energy Technology Data Exchange (ETDEWEB)

Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.

1984-11-01

TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location.

Hydrodynamic Instability, Integrated Code, Laboratory Astrophysics, and Astrophysics

Science.gov (United States)

Takabe, Hideaki

2016-10-01

This is an article for the memorial lecture of Edward Teller Medal and is presented as memorial lecture at the IFSA03 conference held on September 12th, 2003, at Monterey, CA. The author focuses on his main contributions to fusion science and its extension to astrophysics in the field of theory and computation by picking up five topics. The first one is the anomalous resisitivity to hot electrons penetrating over-dense region through the ion wave turbulence driven by the return current compensating the current flow by the hot electrons. It is concluded that almost the same value of potential as the average kinetic energy of the hot electrons is realized to prevent the penetration of the hot electrons. The second is the ablative stabilization of Rayleigh-Taylor instability at ablation front and its dispersion relation so-called Takabe formula. This formula gave a principal guideline for stable target design. The author has developed an integrated code ILESTA (ID & 2D) for analyses and design of laser produced plasma including implosion dynamics. It is also applied to design high gain targets. The third is the development of the integrated code ILESTA. The forth is on Laboratory Astrophysics with intense lasers. This consists of two parts; one is review on its historical background and the other is on how we relate laser plasma to wide-ranging astrophysics and the purposes for promoting such research. In relation to one purpose, I gave a comment on anomalous transport of relativistic electrons in Fast Ignition laser fusion scheme. Finally, I briefly summarize recent activity in relation to application of the author's experience to the development of an integrated code for studying extreme phenomena in astrophysics.

Computation of thermodynamic equilibria of nuclear materials in multi-physics codes

International Nuclear Information System (INIS)

Piro, M.H.; Lewis, B.J.; Thompson, W.T.; Simunovic, S.; Besmann, T.M.

2011-01-01

A new equilibrium thermodynamic solver is being developed with the primary impetus of direct integration into nuclear fuel performance and safety codes to provide improved predictions of fuel behavior. This solver is intended to provide boundary conditions and material properties for continuum transport calculations. There are several legitimate concerns with the use of existing commercial thermodynamic codes: 1) licensing entanglements associated with code distribution, 2) computational performance, and 3) limited capabilities of handling large multi-component systems of interest to the nuclear industry. The development of this solver is specifically aimed at addressing these concerns. In support of this goal, a new numerical algorithm for computing chemical equilibria is presented which is not based on the traditional steepest descent method or 'Gibbs energy minimization' technique. This new approach exploits fundamental principles of equilibrium thermodynamics, which simplifies the optimization equations. The chemical potentials of all species and phases in the system are constrained by the system chemical potentials, and the objective is to minimize the residuals of the mass balance equations. Several numerical advantages are achieved through this simplification, as described in this paper. (author)

Two-dimensional color-code quantum computation

International Nuclear Information System (INIS)

Fowler, Austin G.

2011-01-01

We describe in detail how to perform universal fault-tolerant quantum computation on a two-dimensional color code, making use of only nearest neighbor interactions. Three defects (holes) in the code are used to represent logical qubits. Triple-defect logical qubits are deformed into isolated triangular sections of color code to enable transversal implementation of all single logical qubit Clifford group gates. Controlled-NOT (CNOT) is implemented between pairs of triple-defect logical qubits via braiding.

The archaeology of computer codes - illustrated on the basis of the code SABINE

International Nuclear Information System (INIS)

Sdouz, G.

1987-02-01

Computer codes used by the physics group of the Institute for Reactor Safety are stored on back-up-tapes. However during the last years both the computer and the system have been changed. For new tasks these programmes have to be available. A new procedure is necessary to find and to activate a stored programme. This procedure is illustrated on the basis of the code SABINE. (Author)

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

SKEMA - A computer code to estimate atmospheric dispersion

International Nuclear Information System (INIS)

Sacramento, A.M. do.

1985-01-01

This computer code is a modified version of DWNWND code, developed in Oak Ridge National Laboratory. The Skema code makes an estimative of concentration in air of a material released in atmosphery, by ponctual source. (C.M.) [pt

Computer code qualification program for the Advanced CANDU Reactor

International Nuclear Information System (INIS)

Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.

2003-01-01

Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)

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)

Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

Energy Technology Data Exchange (ETDEWEB)

Young, Michael F.

1999-05-01

The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

Interface design of VSOP'94 computer code for safety analysis

International Nuclear Information System (INIS)

Natsir, Khairina; Andiwijayakusuma, D.; Wahanani, Nursinta Adi; Yazid, Putranto Ilham

2014-01-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects

Interface design of VSOP'94 computer code for safety analysis

Science.gov (United States)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Low Computational Complexity Network Coding For Mobile Networks

DEFF Research Database (Denmark)

Heide, Janus

2012-01-01

Network Coding (NC) is a technique that can provide benefits in many types of networks, some examples from wireless networks are: In relay networks, either the physical or the data link layer, to reduce the number of transmissions. In reliable multicast, to reduce the amount of signaling and enable......-flow coding technique. One of the key challenges of this technique is its inherent computational complexity which can lead to high computational load and energy consumption in particular on the mobile platforms that are the target platform in this work. To increase the coding throughput several...

Computer code conversion using HISTORIAN

International Nuclear Information System (INIS)

Matsumoto, Kiyoshi; Kumakura, Toshimasa.

1990-09-01

When a computer program written for a computer A is converted for a computer B, in general, the A version source program is rewritten for B version. However, in this way of program conversion, the following inconvenient problems arise. 1) The original statements to be rewritten for B version are lost. 2) If the original statements of the A version rewritten for B version would remain as comment lines, the B version source program becomes quite large. 3) When update directives of the program are mailed from the organization which developed the program or when some modifications are needed for the program, it is difficult to point out the part to be updated or modified in the B version source program. To solve these problems, the conversion method using the general-purpose software management aid system, HISTORIAN, has been introduced. This conversion method makes a large computer code a easy-to-use program for use to update, modify or improve after the conversion. This report describes the planning and procedures of the conversion method and the MELPROG-PWR/MOD1 code conversion from the CRAY version to the JAERI FACOM version as an example. This report would provide useful information for those who develop or introduce large programs. (author)

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

Energy Technology Data Exchange (ETDEWEB)

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

2011-02-01

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

Continuous Materiality: Through a Hierarchy of Computational Codes

Directory of Open Access Journals (Sweden)

Jichen Zhu

2008-01-01

Full Text Available The legacy of Cartesian dualism inherent in linguistic theory deeply influences current views on the relation between natural language, computer code, and the physical world. However, the oversimplified distinction between mind and body falls short of capturing the complex interaction between the material and the immaterial. In this paper, we posit a hierarchy of codes to delineate a wide spectrum of continuous materiality. Our research suggests that diagrams in architecture provide a valuable analog for approaching computer code in emergent digital systems. After commenting on ways that Cartesian dualism continues to haunt discussions of code, we turn our attention to diagrams and design morphology. Finally we notice the implications a material understanding of code bears for further research on the relation between human cognition and digital code. Our discussion concludes by noticing several areas that we have projected for ongoing research.

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

International Nuclear Information System (INIS)

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

1997-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-01

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

Holonomic surface codes for fault-tolerant quantum computation

Science.gov (United States)

Zhang, Jiang; Devitt, Simon J.; You, J. Q.; Nori, Franco

2018-02-01

Surface codes can protect quantum information stored in qubits from local errors as long as the per-operation error rate is below a certain threshold. Here we propose holonomic surface codes by harnessing the quantum holonomy of the system. In our scheme, the holonomic gates are built via auxiliary qubits rather than the auxiliary levels in multilevel systems used in conventional holonomic quantum computation. The key advantage of our approach is that the auxiliary qubits are in their ground state before and after each gate operation, so they are not involved in the operation cycles of surface codes. This provides an advantageous way to implement surface codes for fault-tolerant quantum computation.

The FOCON96 1.0 computer code

International Nuclear Information System (INIS)

Merle-Szeremeta, A.; Thomassin, A.

1999-01-01

The Institute of Protection and Nuclear Safety (I.P.S.N.) has developed a computer code, FOCON96 1.0 to calculate the dosimetric consequences of atmospheric radioactive releases from nuclear installations after several years of usual operation. This communication describes the principal characteristics of FOCON96 1.0 and its functionalities. The principal elements of a comparison between FOCON96 1.0 and PC-CREAM ( European computer code developed by the N.R.P.B. and answering the same criteria) are given here. (N.C.)

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

Computer Security: is your code sane?

CERN Multimedia

Stefan Lueders, Computer Security Team

2015-01-01

How many of us write code? Software? Programs? Scripts? How many of us are properly trained in this and how well do we do it? Do we write functional, clean and correct code, without flaws, bugs and vulnerabilities*? In other words: are our codes sane?   Figuring out weaknesses is not that easy (see our quiz in an earlier Bulletin article). Therefore, in order to improve the sanity of your code, prevent common pit-falls, and avoid the bugs and vulnerabilities that can crash your code, or – worse – that can be misused and exploited by attackers, the CERN Computer Security team has reviewed its recommendations for checking the security compliance of your code. “Static Code Analysers” are stand-alone programs that can be run on top of your software stack, regardless of whether it uses Java, C/C++, Perl, PHP, Python, etc. These analysers identify weaknesses and inconsistencies including: employing undeclared variables; expressions resu...

High performance computer code for molecular dynamics simulations

International Nuclear Information System (INIS)

Levay, I.; Toekesi, K.

2007-01-01

Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions

Computer codes for shaping the magnetic field of the JINR phasotron

International Nuclear Information System (INIS)

Zaplatin, N.L.; Morozov, N.A.

1983-01-01

The computer codes providing for the shaping the magnetic field of the JINR high current phasotron are presented. Using these codes the control for the magnetic field mapping was realized in on- or off-line regimes. Then these field parameters were calculated and ferromagnetic correcting elements and trim coils setting were chosen. Some computer codes were realised for the magnetic field horizontal component measurements. The data are presented on some codes possibilities. The codes were used on the EC-1010 and the CDC-6500 computers

Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

Energy Technology Data Exchange (ETDEWEB)

Williams, P. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dickson, T. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yin, S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2007-12-01

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

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

International Nuclear Information System (INIS)

McLaughlin, P.K.

1987-05-01

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

Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

Energy Technology Data Exchange (ETDEWEB)

Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others

1995-12-31

In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.

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

Directory of Open Access Journals (Sweden)

Siniša Šadek

2010-01-01

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

User manual of FRAPCON-I computer code

International Nuclear Information System (INIS)

Chia, C.T.

1985-11-01

The manual for using the FRAPCON-I code implanted by Reactor Department of Brazilian-CNEN to convert IBM FORTRAN in FORTRAN 77 of Honeywell Bull computer is presented. The FRAPCON-I code describes the behaviour of fuel rods of PWR type reactors at stationary state during long periods of burnup. (M.C.K.)

Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers

International Nuclear Information System (INIS)

Woodruff, W.L.

1990-01-01

Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs

Integrated computer-aided design using minicomputers

Science.gov (United States)

Storaasli, O. O.

1980-01-01

Computer-Aided Design/Computer-Aided Manufacturing (CAD/CAM), a highly interactive software, has been implemented on minicomputers at the NASA Langley Research Center. CAD/CAM software integrates many formerly fragmented programs and procedures into one cohesive system; it also includes finite element modeling and analysis, and has been interfaced via a computer network to a relational data base management system and offline plotting devices on mainframe computers. The CAD/CAM software system requires interactive graphics terminals operating at a minimum of 4800 bits/sec transfer rate to a computer. The system is portable and introduces 'interactive graphics', which permits the creation and modification of models interactively. The CAD/CAM system has already produced designs for a large area space platform, a national transonic facility fan blade, and a laminar flow control wind tunnel model. Besides the design/drafting element analysis capability, CAD/CAM provides options to produce an automatic program tooling code to drive a numerically controlled (N/C) machine. Reductions in time for design, engineering, drawing, finite element modeling, and N/C machining will benefit productivity through reduced costs, fewer errors, and a wider range of configuration.

DATING: A computer code for determining allowable temperatures for dry storage of spent fuel in inert and nitrogen gases

International Nuclear Information System (INIS)

Simonen, E.P.; Gilbert, E.R.

1988-12-01

The DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) code can be used to calculate allowable initial temperatures for dry storage of light-water-reactor spent fuel. The calculations are based on the life fraction rule using both measured data and mechanistic equations as reported by Chin et al. (1986). The code is written in FORTRAN and utilizes an efficient numerical integration method for rapid calculations on IBM-compatible personal computers. This report documents the technical basis for the DATING calculations, describes the computational method and code statements, and includes a user's guide with examples. The software for the DATING code is available through the National Energy Software Center operated by Argonne National Laboratory, Argonne, Illinois 60439. 5 refs., 8 figs., 5 tabs

Computation of a BWR Turbine Trip with CATHARE-CRONOS2-FLICA4 Coupled Codes

International Nuclear Information System (INIS)

Mignot, G.; Royer, E.; Rameau, B.; Todorova, N.

2004-01-01

The CEA/DEN modeling and computation results with the CATHARE, CRONOS2, and FLICA4 codes of the Organisation for Economic Co-operation and Development boiling water reactor turbine trip benchmark are presented. The first exercise of the benchmark to model the whole reactor thermal hydraulics with specified power has been performed with the CATHARE system code. Exercise 2, devoted to core thermal-hydraulic neutronic analysis with provided boundary conditions and neutronic cross sections, has been carried out with the CRONOS2 and FLICA4 codes. Finally, exercise 3, combining system thermal hydraulics and core three-dimensional thermal-hydraulics-neutronics, was computed with the three coupled codes: CATHARE, CRONOS2, and FLICA4.Our one-dimensional thermal-hydraulic reactor computation agrees well with the benchmark reference data and demonstrates the capacities of CATHARE to model a turbine trip transient. Coupled three-dimensional thermal-hydraulic and neutronic analysis displays a high sensitivity of the power peak to the core thermal-hydraulic model. The use of at least 100 channels is recommended to achieve reasonable results for integral and local parameters. Deviations between experimental data and exercise 3 results are discussed: timing of events, core pressure drop, and neutronic model. Finally, analysis of extreme scenarios as sensitivity studies on the transient to assess the effect of the scram, the bypass relief valve, and the steam relief valves is presented

Quantum computation with topological codes from qubit to topological fault-tolerance

CERN Document Server

Fujii, Keisuke

2015-01-01

This book presents a self-consistent review of quantum computation with topological quantum codes. The book covers everything required to understand topological fault-tolerant quantum computation, ranging from the definition of the surface code to topological quantum error correction and topological fault-tolerant operations. The underlying basic concepts and powerful tools, such as universal quantum computation, quantum algorithms, stabilizer formalism, and measurement-based quantum computation, are also introduced in a self-consistent way. The interdisciplinary fields between quantum information and other fields of physics such as condensed matter physics and statistical physics are also explored in terms of the topological quantum codes. This book thus provides the first comprehensive description of the whole picture of topological quantum codes and quantum computation with them.

Grid Code Requirements for Wind Power Integration

DEFF Research Database (Denmark)

Wu, Qiuwei

2018-01-01

This chapter reviews the grid code requirements for integration of wind power plants (WPPs). The grid codes reviewed are from the UK, Ireland, Germany, Denmark, Spain, Sweden, the USA, and Canada. Transmission system operators (TSOs) around the world have specified requirements for WPPs under...

A restructuring of the CF/EDF packages for the MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

The CF and EDF packages, which allow the user to define the functions of variables in a database and the usage of an external data file, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To restructure the code, the data transferring methods of the current MELCOR code are modified and then partially adopted into the CF/EDF packages. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as pointers are used to define their addresses. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type without pointers leading to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF/EDF packages addressed in this paper includes a module development and subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code and the trends are almost the same to each other. Therefore the similar approach could be extended to the entire code package for code restructuring. It is expected that the code restructuring will accelerate the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

Statistical Methodologies to Integrate Experimental and Computational Research

Science.gov (United States)

Parker, P. A.; Johnson, R. T.; Montgomery, D. C.

2008-01-01

Development of advanced algorithms for simulating engine flow paths requires the integration of fundamental experiments with the validation of enhanced mathematical models. In this paper, we provide an overview of statistical methods to strategically and efficiently conduct experiments and computational model refinement. Moreover, the integration of experimental and computational research efforts is emphasized. With a statistical engineering perspective, scientific and engineering expertise is combined with statistical sciences to gain deeper insights into experimental phenomenon and code development performance; supporting the overall research objectives. The particular statistical methods discussed are design of experiments, response surface methodology, and uncertainty analysis and planning. Their application is illustrated with a coaxial free jet experiment and a turbulence model refinement investigation. Our goal is to provide an overview, focusing on concepts rather than practice, to demonstrate the benefits of using statistical methods in research and development, thereby encouraging their broader and more systematic application.

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

Structural dynamics in LMFBR containment analysis. A brief survey of computational methods and codes

International Nuclear Information System (INIS)

Chang, Y.W.

1977-01-01

This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers to up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. First, the commonly used finite-difference expressions in the Lagrangian codes will be compared. Sample calculations will be used as a basis for discussing and comparing the accuracy of the various finite-difference representations. The distortion of the meshes will also be compared; the techniques used for eliminating the numerical instabilities will be discussed and compared using examples. Next, the numerical methods used in the Eulerian formulation will be compared, first among themselves and then with the Lagrangian formulations. Special emphasis is placed on the effect of mass diffusion of the Eulerian calculation on the propagation of discontinuities. Implicit and explicit numerical integrations will be discussed and results obtained from these two techniques will be compared. Then, the finite-element methods are compared with the finite-difference methods. The advantages and disadvantages of the two methods will be discussed in detail, together with the versatility and ease of application of the method to containment analysis having complex geometries. It will also be shown that the finite-element equations for a constant-pressure fluid element is identical to the finite-difference equations using contour integrations. Finally, conclusions based on this study will be given

Two-phase computer codes for zero-gravity applications

International Nuclear Information System (INIS)

Krotiuk, W.J.

1986-10-01

This paper discusses the problems existing in the development of computer codes which can analyze the thermal-hydraulic behavior of two-phase fluids especially in low gravity nuclear reactors. The important phenomenon affecting fluid flow and heat transfer in reduced gravity is discussed. The applicability of using existing computer codes for space applications is assessed. Recommendations regarding the use of existing earth based fluid flow and heat transfer correlations are made and deficiencies in these correlations are identified

Interface design of VSOP'94 computer code for safety analysis

Energy Technology Data Exchange (ETDEWEB)

Natsir, Khairina, E-mail: yenny@batan.go.id; Andiwijayakusuma, D.; Wahanani, Nursinta Adi [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia); Yazid, Putranto Ilham [Center for Nuclear Technology, Material and Radiometry- National Nuclear Energy Agency, Jl. Tamansari No.71, Bandung 40132 (Indonesia)

2014-09-30

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Manual for COMSYN: A orbit integration code for the study of beam dynamics in compact synchrotrons

International Nuclear Information System (INIS)

Huang, Y.

1991-10-01

COMSYN is a numerical integration code which is written for the study and design of the compact synchrotrons. An improved 4th-order Runge-Kutta method is used in COMSYN to integrate the exact equations of motion in a rectangular coordinate system. The magnetic field components of the dipole B x , B y and B z can be obtained from either measurement or directly computed data (MAGNUS, TOSCA). A spline interpolation method is then used to get the field value at the particle position. For standard quadrupole and sextupole, the analytical expression is employed to compute its field distribution

Computer codes for designing proton linear accelerators

International Nuclear Information System (INIS)

Kato, Takao

1992-01-01

Computer codes for designing proton linear accelerators are discussed from the viewpoint of not only designing but also construction and operation of the linac. The codes are divided into three categories according to their purposes: 1) design code, 2) generation and simulation code, and 3) electric and magnetic fields calculation code. The role of each category is discussed on the basis of experience at KEK (the design of the 40-MeV proton linac and its construction and operation, and the design of the 1-GeV proton linac). We introduce our recent work relevant to three-dimensional calculation and supercomputer calculation: 1) tuning of MAFIA (three-dimensional electric and magnetic fields calculation code) for supercomputer, 2) examples of three-dimensional calculation of accelerating structures by MAFIA, 3) development of a beam transport code including space charge effects. (author)

IM (Integrity Management) software must show flexibility to local codes

Energy Technology Data Exchange (ETDEWEB)

Brors, Markus [ROSEN Technology and Research Center GmbH (Germany); Diggory, Ian [Macaw Engineering Ltd., Northumberland (United Kingdom)

2009-07-01

There are many internationally recognized codes and standards, such as API 1160 and ASME B31.8S, which help pipeline operators to manage and maintain the integrity of their pipeline networks. However, operators in many countries still use local codes that often reflect the history of pipeline developments in their region and are based on direct experience and research on their pipelines. As pipeline companies come under increasing regulatory and financial pressures to maintain the integrity of their networks, it is important that operators using regional codes are able to benchmark their integrity management schemes against these international standards. Any comprehensive Pipeline Integrity Management System (PIMS) software package should therefore not only incorporate industry standards for pipeline integrity assessment but also be capable of implementing regional codes for comparison purposes. This paper describes the challenges and benefits of incorporating one such set of regional pipeline standards into ROSEN Asset Integrity Management Software (ROAIMS). (author)

Dynamic benchmarking of simulation codes

International Nuclear Information System (INIS)

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

1996-01-01

Computer simulation of nuclear power plant response can be a full-scope control room simulator, an engineering simulator to represent the general behavior of the plant under normal and abnormal conditions, or the modeling of the plant response to conditions that would eventually lead to core damage. In any of these, the underlying foundation for their use in analysing situations, training of vendor/utility personnel, etc. is how well they represent what has been known from industrial experience, large integral experiments and separate effects tests. Typically, simulation codes are benchmarked with some of these; the level of agreement necessary being dependent upon the ultimate use of the simulation tool. However, these analytical models are computer codes, and as a result, the capabilities are continually enhanced, errors are corrected, new situations are imposed on the code that are outside of the original design basis, etc. Consequently, there is a continual need to assure that the benchmarks with important transients are preserved as the computer code evolves. Retention of this benchmarking capability is essential to develop trust in the computer code. Given the evolving world of computer codes, how is this retention of benchmarking capabilities accomplished? For the MAAP4 codes this capability is accomplished through a 'dynamic benchmarking' feature embedded in the source code. In particular, a set of dynamic benchmarks are included in the source code and these are exercised every time the archive codes are upgraded and distributed to the MAAP users. Three different types of dynamic benchmarks are used: plant transients; large integral experiments; and separate effects tests. Each of these is performed in a different manner. The first is accomplished by developing a parameter file for the plant modeled and an input deck to describe the sequence; i.e. the entire MAAP4 code is exercised. The pertinent plant data is included in the source code and the computer

ICAN Computer Code Adapted for Building Materials

Science.gov (United States)

Murthy, Pappu L. N.

1997-01-01

The NASA Lewis Research Center has been involved in developing composite micromechanics and macromechanics theories over the last three decades. These activities have resulted in several composite mechanics theories and structural analysis codes whose applications range from material behavior design and analysis to structural component response. One of these computer codes, the Integrated Composite Analyzer (ICAN), is designed primarily to address issues related to designing polymer matrix composites and predicting their properties - including hygral, thermal, and mechanical load effects. Recently, under a cost-sharing cooperative agreement with a Fortune 500 corporation, Master Builders Inc., ICAN was adapted to analyze building materials. The high costs and technical difficulties involved with the fabrication of continuous-fiber-reinforced composites sometimes limit their use. Particulate-reinforced composites can be thought of as a viable alternative. They are as easily processed to near-net shape as monolithic materials, yet have the improved stiffness, strength, and fracture toughness that is characteristic of continuous-fiber-reinforced composites. For example, particlereinforced metal-matrix composites show great potential for a variety of automotive applications, such as disk brake rotors, connecting rods, cylinder liners, and other hightemperature applications. Building materials, such as concrete, can be thought of as one of the oldest materials in this category of multiphase, particle-reinforced materials. The adaptation of ICAN to analyze particle-reinforced composite materials involved the development of new micromechanics-based theories. A derivative of the ICAN code, ICAN/PART, was developed and delivered to Master Builders Inc. as a part of the cooperative activity.

A compendium of computer codes in fault tree analysis

International Nuclear Information System (INIS)

Lydell, B.

1981-03-01

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

Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)

International Nuclear Information System (INIS)

Maiorino, J.R.

1990-01-01

The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)

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)

Independent peer review of nuclear safety computer codes

International Nuclear Information System (INIS)

Boyack, B.E.; Jenks, R.P.

1993-01-01

A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC

Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

1988-01-01

This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)

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

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

Radiological impact assessment in Malaysia using RESRAD computer code

International Nuclear Information System (INIS)

Syed Hakimi Sakuma Syed Ahmad; Khairuddin Mohamad Kontol; Razali Hamzah

1999-01-01

Radiological Impact Assessment (RIA) can be conducted in Malaysia by using the RESRAD computer code developed by Argonne National Laboratory, U.S.A. The code can do analysis to derive site specific guidelines for allowable residual concentrations of radionuclides in soil. Concepts of the RIA in the context of waste management concern in Malaysia, some regulatory information and assess status of data collection are shown. Appropriate use scenarios and site specific parameters are used as much as possible so as to be realistic so that will reasonably ensure that individual dose limits and or constraints will be achieved. Case study have been conducted to fulfil Atomic Energy Licensing Board (AELB) requirements where for disposal purpose the operator must be required to carry out. a radiological impact assessment to all proposed disposals. This is to demonstrate that no member of public will be exposed to more than 1 mSv/year from all activities. Results obtained from analyses show the RESRAD computer code is able to calculate doses, risks, and guideline values. Sensitivity analysis by the computer code shows that the parameters used as input are justified so as to improve confidence to the public and the AELB the results of the analysis. The computer code can also be used as an initial assessment to conduct screening assessment in order to determine a proper disposal site. (Author)

CBP TOOLBOX VERSION 2.0: CODE INTEGRATION ENHANCEMENTS

Energy Technology Data Exchange (ETDEWEB)

Smith, F.; Flach, G.; BROWN, K.

2013-06-01

This report describes enhancements made to code integration aspects of the Cementitious Barriers Project (CBP) Toolbox as a result of development work performed at the Savannah River National Laboratory (SRNL) in collaboration with Vanderbilt University (VU) in the first half of fiscal year 2013. Code integration refers to the interfacing to standalone CBP partner codes, used to analyze the performance of cementitious materials, with the CBP Software Toolbox. The most significant enhancements are: 1) Improved graphical display of model results. 2) Improved error analysis and reporting. 3) Increase in the default maximum model mesh size from 301 to 501 nodes. 4) The ability to set the LeachXS/Orchestra simulation times through the GoldSim interface. These code interface enhancements have been included in a new release (Version 2.0) of the CBP Toolbox.

A DOE Computer Code Toolbox: Issues and Opportunities

International Nuclear Information System (INIS)

Vincent, A.M. III

2001-01-01

The initial activities of a Department of Energy (DOE) Safety Analysis Software Group to establish a Safety Analysis Toolbox of computer models are discussed. The toolbox shall be a DOE Complex repository of verified and validated computer models that are configuration-controlled and made available for specific accident analysis applications. The toolbox concept was recommended by the Defense Nuclear Facilities Safety Board staff as a mechanism to partially address Software Quality Assurance issues. Toolbox candidate codes have been identified through review of a DOE Survey of Software practices and processes, and through consideration of earlier findings of the Accident Phenomenology and Consequence Evaluation program sponsored by the DOE National Nuclear Security Agency/Office of Defense Programs. Planning is described to collect these high-use codes, apply tailored SQA specific to the individual codes, and implement the software toolbox concept. While issues exist such as resource allocation and the interface among code developers, code users, and toolbox maintainers, significant benefits can be achieved through a centralized toolbox and subsequent standardized applications

Parallel computing by Monte Carlo codes MVP/GMVP

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Nakagawa, Masayuki; Mori, Takamasa

2001-01-01

General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel paragon and a distributed-memory scalar-parallel computer Hitachi SR2201, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)

Utilization of KENO-IV computer code with HANSEN-ROACH library

International Nuclear Information System (INIS)

Lima Barros, M. de; Vellozo, S.O.

1982-01-01

Several analysis with KENO-IV computer code, which is based in the Monte Carlo method, and the cross section library HANSEN-ROACH, were done, aiming to present the more convenient form to execute criticality calculations with this computer code and this cross sections. (E.G.) [pt

Heat Transfer treatment in computer codes for safety analysis

International Nuclear Information System (INIS)

Jerele, A.; Gregoric, M.

1984-01-01

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

FLASH: A finite element computer code for variably saturated flow

International Nuclear Information System (INIS)

Baca, R.G.; Magnuson, S.O.

1992-05-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model, referred to as the FLASH computer code, is designed to simulate two-dimensional fluid flow in fractured-porous media. The code is specifically designed to model variably saturated flow in an arid site vadose zone and saturated flow in an unconfined aquifer. In addition, the code also has the capability to simulate heat conduction in the vadose zone. This report presents the following: description of the conceptual frame-work and mathematical theory; derivations of the finite element techniques and algorithms; computational examples that illustrate the capability of the code; and input instructions for the general use of the code. The FLASH computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of Energy Order 5820.2A

Computing the Feng-Rao distances for codes from order domains

DEFF Research Database (Denmark)

Ruano Benito, Diego

2007-01-01

We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis.......We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis....

STARS: An Integrated, Multidisciplinary, Finite-Element, Structural, Fluids, Aeroelastic, and Aeroservoelastic Analysis Computer Program

Science.gov (United States)

Gupta, K. K.

1997-01-01

A multidisciplinary, finite element-based, highly graphics-oriented, linear and nonlinear analysis capability that includes such disciplines as structures, heat transfer, linear aerodynamics, computational fluid dynamics, and controls engineering has been achieved by integrating several new modules in the original STARS (STructural Analysis RoutineS) computer program. Each individual analysis module is general-purpose in nature and is effectively integrated to yield aeroelastic and aeroservoelastic solutions of complex engineering problems. Examples of advanced NASA Dryden Flight Research Center projects analyzed by the code in recent years include the X-29A, F-18 High Alpha Research Vehicle/Thrust Vectoring Control System, B-52/Pegasus Generic Hypersonics, National AeroSpace Plane (NASP), SR-71/Hypersonic Launch Vehicle, and High Speed Civil Transport (HSCT) projects. Extensive graphics capabilities exist for convenient model development and postprocessing of analysis results. The program is written in modular form in standard FORTRAN language to run on a variety of computers, such as the IBM RISC/6000, SGI, DEC, Cray, and personal computer; associated graphics codes use OpenGL and IBM/graPHIGS language for color depiction. This program is available from COSMIC, the NASA agency for distribution of computer programs.

An Analysis of Countries which have Integrated Coding into their Curricula and the Content Analysis of Academic Studies on Coding Training in Turkey

Directory of Open Access Journals (Sweden)

Hüseyin Uzunboylu

2017-11-01

Full Text Available The first aim is to conduct a general analysis of countries which have integrated coding training into their curricula, and the second aim is to conduct a content analysis of studies on coding training in Turkey. It was identified that there are only a few academic studies on coding training in Turkey, and that the majority of them were published in 2016, the intended population was mainly “undergraduate students” and that the majority of these students were Computer Education and Instructional Technology undergraduates. It was determined that the studies mainly focused on the subjects of “programming” and “Scratch”, the terms programming and coding were used as synonyms, most of the studies were carried out using quantitative methods and data was obtained mostly by literature review and scale/survey interval techniques.

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

International Nuclear Information System (INIS)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

2011-01-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

Energy Technology Data Exchange (ETDEWEB)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

2011-03-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

Control rod computer code IAMCOS: general theory and numerical methods

International Nuclear Information System (INIS)

West, G.

1982-11-01

IAMCOS is a computer code for the description of mechanical and thermal behavior of cylindrical control rods for fast breeders. This code version was applied, tested and modified from 1979 to 1981. In this report are described the basic model (02 version), theoretical definitions and computation methods [fr

Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding

International Nuclear Information System (INIS)

Alexander D Efanov; Vladimir N Vinogradov; Victor V Sergeev; Oleg A Sudnitsyn

2005-01-01

Full text of publication follows: The computational assessment of reactor core components behavior under accident conditions is impossible without knowledge of the thermal-hydraulic processes occurring in this case. The adequacy of the results obtained using the computer codes to the real processes is verified by carrying out a number of standard problems. In 2000-2003, the fulfillment of three Russian standard problems on WWER core reflooding was arranged using the experiments on full-height electrically heated WWER 37-rod bundle model cooldown in regimes of bottom (SP-1), top (SP-2) and combined (SP-3) reflooding. The representatives from the eight MINATOM's organizations took part in this work, in the course of which the 'blind' and posttest calculations were performed using various versions of the RELAP5, ATHLET, CATHARE, COBRA-TF, TRAP, KORSAR computer codes. The paper presents a brief description of the test facility, test section, test scenarios and conditions as well as the basic results of computational analysis of the experiments. The analysis of the test data revealed a significantly non-one-dimensional nature of cooldown and rewetting of heater rods heated up to a high temperature in a model bundle. This was most pronounced at top and combined reflooding. The verification of the model reflooding computer codes showed that most of computer codes fairly predict the peak rod temperature and the time of bundle cooldown. The exception is provided by the results of calculations with the ATHLET and CATHARE codes. The nature and rate of rewetting front advance in the lower half of the bundle are fairly predicted practically by all computer codes. The disagreement between the calculations and experimental results for the upper half of the bundle is caused by the difficulties of computational simulation of multidimensional effects by 1-D computer codes. In this regard, a quasi-two-dimensional computer code COBRA-TF offers certain advantages. Overall, the closest

Thermohydraulic analysis of nuclear power plant accidents by computer codes

International Nuclear Information System (INIS)

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

1982-01-01

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

The SEDA computer code and its utilization for Angra 1

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1988-11-01

The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the environment. This code is now available for an IBM PC-XT. The computer environment, the files used, data, the programining structure and the models used are presented. The input data and results for two sample case are described. (author) [pt

Interface between computational fluid dynamics (CFD) and plant analysis computer codes

International Nuclear Information System (INIS)

Coffield, R.D.; Dunckhorst, F.F.; Tomlinson, E.T.; Welch, J.W.

1993-01-01

Computational fluid dynamics (CFD) can provide valuable input to the development of advanced plant analysis computer codes. The types of interfacing discussed in this paper will directly contribute to modeling and accuracy improvements throughout the plant system and should result in significant reduction of design conservatisms that have been applied to such analyses in the past

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.

Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †.

Science.gov (United States)

Murdani, Muhammad Harist; Kwon, Joonho; Choi, Yoon-Ho; Hong, Bonghee

2018-03-24

In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes ( Ad-Hoc ) and neighborhood proximity ( Top-K ). Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.

Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †

Directory of Open Access Journals (Sweden)

Muhammad Harist Murdani

2018-03-01

Full Text Available In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes (Ad-Hoc and neighborhood proximity (Top-K. Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.

Integrated Tiger Series of electron/photon Monte Carlo transport codes: a user's guide for use on IBM mainframes

International Nuclear Information System (INIS)

Kirk, B.L.

1985-12-01

The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler

Single integrated device for optical CDMA code processing in dual-code environment.

Science.gov (United States)

Huang, Yue-Kai; Glesk, Ivan; Greiner, Christoph M; Iazkov, Dmitri; Mossberg, Thomas W; Wang, Ting; Prucnal, Paul R

2007-06-11

We report on the design, fabrication and performance of a matching integrated optical CDMA encoder-decoder pair based on holographic Bragg reflector technology. Simultaneous encoding/decoding operation of two multiple wavelength-hopping time-spreading codes was successfully demonstrated and shown to support two error-free OCDMA links at OC-24. A double-pass scheme was employed in the devices to enable the use of longer code length.

TRAC-PF1/MOD1 computer code

International Nuclear Information System (INIS)

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

1984-01-01

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

Case studies in Gaussian process modelling of computer codes

International Nuclear Information System (INIS)

Kennedy, Marc C.; Anderson, Clive W.; Conti, Stefano; O'Hagan, Anthony

2006-01-01

In this paper we present a number of recent applications in which an emulator of a computer code is created using a Gaussian process model. Tools are then applied to the emulator to perform sensitivity analysis and uncertainty analysis. Sensitivity analysis is used both as an aid to model improvement and as a guide to how much the output uncertainty might be reduced by learning about specific inputs. Uncertainty analysis allows us to reflect output uncertainty due to unknown input parameters, when the finished code is used for prediction. The computer codes themselves are currently being developed within the UK Centre for Terrestrial Carbon Dynamics

Computer codes for beam dynamics analysis of cyclotronlike accelerators

Science.gov (United States)

Smirnov, V.

2017-12-01

Computer codes suitable for the study of beam dynamics in cyclotronlike (classical and isochronous cyclotrons, synchrocyclotrons, and fixed field alternating gradient) accelerators are reviewed. Computer modeling of cyclotron segments, such as the central zone, acceleration region, and extraction system is considered. The author does not claim to give a full and detailed description of the methods and algorithms used in the codes. Special attention is paid to the codes already proven and confirmed at the existing accelerating facilities. The description of the programs prepared in the worldwide known accelerator centers is provided. The basic features of the programs available to users and limitations of their applicability are described.

LATTICE: an interactive lattice computer code

International Nuclear Information System (INIS)

Staples, J.

1976-10-01

LATTICE is a computer code which enables an interactive user to calculate the functions of a synchrotron lattice. This program satisfies the requirements at LBL for a simple interactive lattice program by borrowing ideas from both TRANSPORT and SYNCH. A fitting routine is included

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

International Nuclear Information System (INIS)

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

1997-09-01

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

User's manual for computer code RIBD-II, a fission product inventory code

International Nuclear Information System (INIS)

Marr, D.R.

1975-01-01

The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

SWIMS: a small-angle multiple scattering computer code

International Nuclear Information System (INIS)

Sayer, R.O.

1976-07-01

SWIMS (Sigmund and WInterbon Multiple Scattering) is a computer code for calculation of the angular dispersion of ion beams that undergo small-angle, incoherent multiple scattering by gaseous or solid media. The code uses the tabulated angular distributions of Sigmund and Winterbon for a Thomas-Fermi screened Coulomb potential. The fraction of the incident beam scattered into a cone defined by the polar angle α is computed as a function of α for reduced thicknesses over the range 0.01 less than or equal to tau less than or equal to 10.0. 1 figure, 2 tables

War of ontology worlds: mathematics, computer code, or Esperanto?

Science.gov (United States)

Rzhetsky, Andrey; Evans, James A

2011-09-01

The use of structured knowledge representations-ontologies and terminologies-has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies.

SIMIFR: A code to simulate material movement in the Integral Fast Reactor

International Nuclear Information System (INIS)

White, A.M.; Orechwa, Yuri.

1991-01-01

The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

International Nuclear Information System (INIS)

Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

1997-12-01

FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes' integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6)

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

International Nuclear Information System (INIS)

1975-06-01

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

User Instructions for the Systems Assessment Capability, Rev. 0, Computer Codes Volume 2: Impact Modules

International Nuclear Information System (INIS)

Eslinger, Paul W.; Arimescu, Carmen; Kanyid, Beverly A.; Miley, Terri B.

2001-01-01

One activity of the Department of Energy?s Groundwater/Vadose Zone Integration Project is an assessment of cumulative impacts from Hanford Site wastes on the subsurface environment and the Columbia River. Through the application of a system assessment capability (SAC), decisions for each cleanup and disposal action will be able to take into account the composite effect of other cleanup and disposal actions. The SAC has developed a suite of computer programs to simulate the migration of contaminants (analytes) present on the Hanford Site and to assess the potential impacts of the analytes, including dose to humans, socio-cultural impacts, economic impacts, and ecological impacts. The general approach to handling uncertainty in the SAC computer codes is a Monte Carlo approach. Conceptually, one generates a value for every stochastic parameter in the code (the entire sequence of modules from inventory through transport and impacts) and then executes the simulation, obtaining an output value, or result. This document provides user instructions for the SAC codes that generate human, ecological, economic, and cultural impacts

User's manual for the NEFTRAN II computer code

International Nuclear Information System (INIS)

Olague, N.E.; Campbell, J.E.; Leigh, C.D.; Longsine, D.E.

1991-02-01

This document describes the NEFTRAN II (NEtwork Flow and TRANsport in Time-Dependent Velocity Fields) computer code and is intended to provide the reader with sufficient information to use the code. NEFTRAN II was developed as part of a performance assessment methodology for storage of high-level nuclear waste in unsaturated, welded tuff. NEFTRAN II is a successor to the NEFTRAN and NWFT/DVM computer codes and contains several new capabilities. These capabilities include: (1) the ability to input pore velocities directly to the transport model and bypass the network fluid flow model, (2) the ability to transport radionuclides in time-dependent velocity fields, (3) the ability to account for the effect of time-dependent saturation changes on the retardation factor, and (4) the ability to account for time-dependent flow rates through the source regime. In addition to these changes, the input to NEFTRAN II has been modified to be more convenient for the user. This document is divided into four main sections consisting of (1) a description of all the models contained in the code, (2) a description of the program and subprograms in the code, (3) a data input guide and (4) verification and sample problems. Although NEFTRAN II is the fourth generation code, this document is a complete description of the code and reference to past user's manuals should not be necessary. 19 refs., 33 figs., 25 tabs

Gender codes why women are leaving computing

CERN Document Server

Misa, Thomas J

2010-01-01

The computing profession is facing a serious gender crisis. Women are abandoning the computing field at an alarming rate. Fewer are entering the profession than anytime in the past twenty-five years, while too many are leaving the field in mid-career. With a maximum of insight and a minimum of jargon, Gender Codes explains the complex social and cultural processes at work in gender and computing today. Edited by Thomas Misa and featuring a Foreword by Linda Shafer, Chair of the IEEE Computer Society Press, this insightful collection of essays explores the persisting gender imbalance in computing and presents a clear course of action for turning things around.

A restructuring of the MELCOR fission product packages for the MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

The RN1/RN2 packages, which are the fission product-related packages in MELCOR, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the RN1/RN2 package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1/RN2 package addressed in this paper includes a module development, subroutine modification, and the treatment of MELGEN, which generates the data file, as well as MELCOR, which is processing the calculation. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerate the code domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

Statistical screening of input variables in a complex computer code

International Nuclear Information System (INIS)

Krieger, T.J.

1982-01-01

A method is presented for ''statistical screening'' of input variables in a complex computer code. The object is to determine the ''effective'' or important input variables by estimating the relative magnitudes of their associated sensitivity coefficients. This is accomplished by performing a numerical experiment consisting of a relatively small number of computer runs with the code followed by a statistical analysis of the results. A formula for estimating the sensitivity coefficients is derived. Reference is made to an earlier work in which the method was applied to a complex reactor code with good results

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

International Nuclear Information System (INIS)

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

1976-05-01

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

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)

Citham-2 computer code-User manual

International Nuclear Information System (INIS)

Batista, J.L.

1984-01-01

The procedures and the input data for the Citham-2 computer code are described. It is a subroutine that modifies the nuclide concentration taking in account its burn and prepares cross sections library in 2,3 or 4 energy groups, to the used for Citation program. (E.G.) [pt

Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1984-12-01

The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

Computer code for quantitative ALARA evaluations

International Nuclear Information System (INIS)

Voilleque, P.G.

1984-01-01

A FORTRAN computer code has been developed to simplify the determination of whether dose reduction actions meet the as low as is reasonably achievable (ALARA) criterion. The calculations are based on the methodology developed for the Atomic Industrial Forum. The code is used for analyses of eight types of dose reduction actions, characterized as follows: reduce dose rate, reduce job frequency, reduce productive working time, reduce crew size, increase administrative dose limit for the task, and increase the workers' time utilization and dose utilization through (a) improved working conditions, (b) basic skill training, or (c) refresher training for special skills. For each type of action, two analysis modes are available. The first is a generic analysis in which the program computes potential benefits (in dollars) for a range of possible improvements, e.g., for a range of lower dose rates. Generic analyses are most useful in the planning stage and for evaluating the general feasibility of alternative approaches. The second is a specific analysis in which the potential annual benefits of a specific level of improvement and the annual implementation cost are compared. The potential benefits reflect savings in operational and societal costs that can be realized if occupational radiation doses are reduced. Because the potential benefits depend upon many variables which characterize the job, the workplace, and the workers, there is no unique relationship between the potential dollar savings and the dose savings. The computer code permits rapid quantitative analyses of alternatives and is a tool that supplements the health physicist's professional judgment. The program output provides a rational basis for decision-making and a record of the assumptions employed

Computer-aided software understanding systems to enhance confidence of scientific codes

International Nuclear Information System (INIS)

Sheng, G.; Oeren, T.I.

1991-01-01

A unique characteristic of nuclear waste disposal is the very long time span over which the combined engineered and natural containment system must remain effective: hundreds of thousands of years. Since there is no precedent in human history for such an endeavour, simulation with the use of computers is the only means we have of forecasting possible future outcomes quantitatively. The need for reliable models and software to make such forecasts so far into the future is obvious. One of the critical elements necessary to ensure reliability is the degree of reviewability of the computer program. Among others, there are two very important reasons for this. Firstly, if there is to be any chance at all of validating the conceptual models as implemented by the computer code, peer reviewers must be able to see and understand what the program is doing. It is all but impossible to achieve this understanding by just looking at the code due to possible unfamiliarity with the language and often due as well to the length and complexity of the code. Secondly, a thorough understanding of the code is also necessary to carry out code maintenance activities which include among others, error detection, error correction and code modification for purposes of enhancing its performance, functionality or to adapt it to a changed environment. The emerging concepts of computer-aided software understanding and reverse engineering can answer precisely these needs. This paper will discuss the role they can play in enhancing the confidence one has on computer codes and several examples will be provided. Finally a brief discussion of combining state-of-art forward engineering systems with reverse engineering systems will show how powerfully they can contribute to the overall quality assurance of a computer program. (13 refs., 7 figs.)

MELMRK 2.0: A description of computer models and results of code testing

International Nuclear Information System (INIS)

Wittman, R.S.; Denny, V.; Mertol, A.

1992-01-01

An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly

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

International Nuclear Information System (INIS)

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

1991-02-01

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

Independent validation testing of the FLAME computer code, Version 1.0

International Nuclear Information System (INIS)

Martian, P.; Chung, J.N.

1992-07-01

Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions

Development and using computer codes for improvement of defect assembly detection on Russian WWER NPPs

International Nuclear Information System (INIS)

Likhanskii, V.; Evdokimov, I.; Zborovskii, V.; Kanukova, V.; Sorokin, A.; Taran, M.; Ugrumov, A.; Riabinin, Y.

2009-01-01

Diagnostic methods of fuel failure detection for improving the radiation safety and shortening of fuel reload time at Russian WWERs are currently in development . The works include creation new computer means for increase of effectiveness of fuel monitoring and reliability of leakage tests. Reliability of failure detection can be noticeably improved when we apply an integrated approach including the following methods. The first is fuel failure analysis under operating conditions. Analysis is performed with the pilot version of the expert system, which has been developed on the basis of the mechanistic code RTOP-CA. The second stage of failure monitoring is 'sipping' tests in the mast of the refueling machine. The leakage tests are the final stage of failure monitoring. A new technique with pressure cycling in the specialized casks was introduced to meet the requirements of higher reliability in detection/confirmation of the leakages. Measurements of the activity release kinetics during the pressure cycling and handling of the acquired data with the RTOP-LT code enable to evaluate a defect size in leaking fuel assembly. The mechanistic codes RTOP-CA and RTOP-LT were verified on a base of specialized experimental data and currently the code were certified by Russian authorities Rostechnadzor. Now the pressure cycling method in the specialized casks has official status and is utilized at the all Russian WWER units. Some results of application of the integrated approach to fuel failure monitoring at several Russian NPPs with WWER units are reported in the present paper. Predictions of the current version of the expert system are compared with the results of the leakage tests and with the estimations of the defect size by the pressure cycling technique. Using the RTOP-CA code the level of activity is assessed for the following fuel campaign if the leaking fuel assembly was decided to be reloaded into the core. A project of the automated computer system on the basis of

Compilation of the abstracts of nuclear computer codes available at CPD/IPEN

International Nuclear Information System (INIS)

Granzotto, A.; Gouveia, A.S. de; Lourencao, E.M.

1981-06-01

A compilation of all computer codes available at IPEN in S.Paulo are presented. These computer codes are classified according to Argonne National Laboratory - and Energy Nuclear Agency schedule. (E.G.) [pt

Development and Verification of the Computer Codes for the Fast Reactors Nuclear Safety Justification

International Nuclear Information System (INIS)

Kisselev, A.E.; Mosunova, N.A.; Strizhov, V.F.

2015-01-01

The information on the status of the work on development of the system of the nuclear safety codes for fast liquid metal reactors is presented in paper. The purpose of the work is to create an instrument for NPP neutronic, thermohydraulic and strength justification including human and environment radiation safety. The main task that is to be solved by the system of codes developed is the analysis of the broad spectrum of phenomena taking place on the NPP (including reactor itself, NPP components, containment rooms, industrial site and surrounding area) and analysis of the impact of the regular and accidental releases on the environment. The code system is oriented on the ability of fully integrated modeling of the NPP behavior in the coupled definition accounting for the wide range of significant phenomena taking place on the NPP under normal and accident conditions. It is based on the models that meet the state-of-the-art knowledge level. The codes incorporate advanced numerical methods and modern programming technologies oriented on the high-performance computing systems. The information on the status of the work on verification of the separate codes of the system of codes is also presented. (author)

The computer code EURDYN-1M (release 2). User's manual

International Nuclear Information System (INIS)

1982-01-01

EURDYN-1M is a finite element computer code developed at J.R.C. Ispra to compute the response of two-dimensional coupled fluid-structure configurations to transient dynamic loading for reactor safety studies. This report gives instructions for preparing input data to EURDYN-1M, release 2, and describes a test problem in order to illustrate both the input and the output of the code

Additional extensions to the NASCAP computer code, volume 3

Science.gov (United States)

Mandell, M. J.; Cooke, D. L.

1981-01-01

The ION computer code is designed to calculate charge exchange ion densities, electric potentials, plasma temperatures, and current densities external to a neutralized ion engine in R-Z geometry. The present version assumes the beam ion current and density to be known and specified, and the neutralizing electrons to originate from a hot-wire ring surrounding the beam orifice. The plasma is treated as being resistive, with an electron relaxation time comparable to the plasma frequency. Together with the thermal and electrical boundary conditions described below and other straightforward engine parameters, these assumptions suffice to determine the required quantities. The ION code, written in ASCII FORTRAN for UNIVAC 1100 series computers, is designed to be run interactively, although it can also be run in batch mode. The input is free-format, and the output is mainly graphical, using the machine-independent graphics developed for the NASCAP code. The executive routine calls the code's major subroutines in user-specified order, and the code allows great latitude for restart and parameter change.

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)

Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

International Nuclear Information System (INIS)

Shimooke, Takanori; Yoshida, Kazuo

1979-05-01

The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

GASFLOW computer code (physical models and input data)

International Nuclear Information System (INIS)

Muehlbauer, Petr

2007-11-01

The GASFLOW computer code was developed jointly by the Los Alamos National Laboratory, USA, and Forschungszentrum Karlsruhe, Germany. The code is primarily intended for calculations of the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and in other facilities. The physical models and the input data are described, and a commented simple calculation is presented

A zero-dimensional EXTRAP computer code

International Nuclear Information System (INIS)

Karlsson, P.

1982-10-01

A zero-dimensional computer code has been designed for the EXTRAP experiment to predict the density and the temperature and their dependence upon paramenters such as the plasma current and the filling pressure of neutral gas. EXTRAP is a Z-pinch immersed in a vacuum octupole field and could be either linear or toroidal. In this code the density and temperature are assumed to be constant from the axis up to a breaking point from where they decrease linearly in the radial direction out to the plasma radius. All quantities, however, are averaged over the plasma volume thus giving the zero-dimensional character of the code. The particle, momentum and energy one-fluid equations are solved including the effects of the surrounding neutral gas and oxygen impurities. The code shows that the temperature and density are very sensitive to the shape of the plasma, flatter profiles giving higher temperatures and densities. The temperature, however, is not strongly affected for oxygen concentration less than 2% and is well above the radiation barrier even for higher concentrations. (Author)

The implementation of CP1 computer code in the Honeywell Bull computer in Brazilian Nuclear Energy Commission (CNEN)

International Nuclear Information System (INIS)

Couto, R.T.

1987-01-01

The implementation of the CP1 computer code in the Honeywell Bull computer in Brazilian Nuclear Energy Comission is presented. CP1 is a computer code used to solve the equations of punctual kinetic with Doppler feed back from the system temperature variation based on the Newton refrigeration equation (E.G.) [pt

Interface code between WIMS-AECL and RFSP-IST for coupling computing

International Nuclear Information System (INIS)

Xu Liangwang; Liu Yu; Jia Baoshan

2007-01-01

A code based on the protocols of Telnet and FTP is developed with C++ for coupling computing between WIMS-AECL and RFSP-IST. the input document of WIMS-AECL and RFSP-ISP cna be generated automatically and be submitted to server, the output document will be downloaded by the end of computing. the function of analyzing standard output document is also included in this code. After simple updating, this code can meet the requirement of other code using input document, e.g. CATHENA. A pilot study of the relation between void fraction and reactivity in TACR, some valuable conclusions has been achieved. (authors)

Committed to the Honor Code: An Investment Model Analysis of Academic Integrity

Science.gov (United States)

Dix, Emily L.; Emery, Lydia F.; Le, Benjamin

2014-01-01

Educators worldwide face challenges surrounding academic integrity. The development of honor codes can promote academic integrity, but understanding how and why honor codes affect behavior is critical to their successful implementation. To date, research has not examined how students' "relationship" to an honor code predicts…

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

International Nuclear Information System (INIS)

Dickens, J.K.

1981-03-01

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

Numerical simulations of inertial confinement fusion hohlraum with LARED-integration code

International Nuclear Information System (INIS)

Li Jinghong; Li Shuanggui; Zhai Chuanlei

2011-01-01

In the target design of the Inertial Confinement Fusion (ICF) program, it is common practice to apply radiation hydrodynamics code to study the key physical processes happened in ICF process, such as hohlraum physics, radiation drive symmetry, capsule implosion physics in the radiation-drive approach of ICF. Recently, many efforts have been done to develop our 2D integrated simulation capability of laser fusion with a variety of optional physical models and numerical methods. In order to effectively integrate the existing codes and to facilitate the development of new codes, we are developing an object-oriented structured-mesh parallel code-supporting infrastructure, called JASMIN. Based on two-dimensional three-temperature hohlraum physics code LARED-H and two-dimensional multi-group radiative transfer code LARED-R, we develop a new generation two-dimensional laser fusion code under the JASMIN infrastructure, which enable us to simulate the whole process of laser fusion from the laser beams' entrance into the hohlraum to the end of implosion. In this paper, we will give a brief description of our new-generation two-dimensional laser fusion code, named LARED-Integration, especially in its physical models, and present some simulation results of holhraum. (author)

Computational Modelling of the Structural Integrity following Mass-Loss in Polymeric Charred Cellular Solids

OpenAIRE

J. P. M. Whitty; J. Francis; J. Howe; B. Henderson

2014-01-01

A novel computational technique is presented for embedding mass-loss due to burning into the ANSYS finite element modelling code. The approaches employ a range of computational modelling methods in order to provide more complete theoretical treatment of thermoelasticity absent from the literature for over six decades. Techniques are employed to evaluate structural integrity (namely, elastic moduli, Poisson’s ratios, and compressive brittle strength) of honeycomb systems known to approximate t...

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)

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)

H0 precessor computer code

International Nuclear Information System (INIS)

van Dyck, O.B.; Floyd, R.A.

1981-05-01

A spin precessor using H - to H 0 stripping, followed by small precession magnets, has been developed for the LAMPF 800-MeV polarized H - beam. The performance of the system was studied with the computer code documented in this report. The report starts from the fundamental physics of a system of spins with hyperfine coupling in a magnetic field and contains many examples of beam behavior as calculated by the program

Computation of the Genetic Code

Science.gov (United States)

Kozlov, Nicolay N.; Kozlova, Olga N.

2018-03-01

One of the problems in the development of mathematical theory of the genetic code (summary is presented in [1], the detailed -to [2]) is the problem of the calculation of the genetic code. Similar problems in the world is unknown and could be delivered only in the 21st century. One approach to solving this problem is devoted to this work. For the first time provides a detailed description of the method of calculation of the genetic code, the idea of which was first published earlier [3]), and the choice of one of the most important sets for the calculation was based on an article [4]. Such a set of amino acid corresponds to a complete set of representations of the plurality of overlapping triple gene belonging to the same DNA strand. A separate issue was the initial point, triggering an iterative search process all codes submitted by the initial data. Mathematical analysis has shown that the said set contains some ambiguities, which have been founded because of our proposed compressed representation of the set. As a result, the developed method of calculation was limited to the two main stages of research, where the first stage only the of the area were used in the calculations. The proposed approach will significantly reduce the amount of computations at each step in this complex discrete structure.

PORPST: A statistical postprocessor for the PORMC computer code

International Nuclear Information System (INIS)

Eslinger, P.W.; Didier, B.T.

1991-06-01

This report describes the theory underlying the PORPST code and gives details for using the code. The PORPST code is designed to do statistical postprocessing on files written by the PORMC computer code. The data written by PORMC are summarized in terms of means, variances, standard deviations, or statistical distributions. In addition, the PORPST code provides for plotting of the results, either internal to the code or through use of the CONTOUR3 postprocessor. Section 2.0 discusses the mathematical basis of the code, and Section 3.0 discusses the code structure. Section 4.0 describes the free-format point command language. Section 5.0 describes in detail the commands to run the program. Section 6.0 provides an example program run, and Section 7.0 provides the references. 11 refs., 1 fig., 17 tabs

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.

Microdosimetry computation code of internal sources - MICRODOSE 1

International Nuclear Information System (INIS)

Li Weibo; Zheng Wenzhong; Ye Changqing

1995-01-01

This paper describes a microdosimetry computation code, MICRODOSE 1, on the basis of the following described methods: (1) the method of calculating f 1 (z) for charged particle in the unit density tissues; (2) the method of calculating f(z) for a point source; (3) the method of applying the Fourier transform theory to the calculation of the compound Poisson process; (4) the method of using fast Fourier transform technique to determine f(z) and, giving some computed examples based on the code, MICRODOSE 1, including alpha particles emitted from 239 Pu in the alveolar lung tissues and from radon progeny RaA and RAC in the human respiratory tract. (author). 13 refs., 6 figs

Probabilistic evaluations for CANTUP computer code analysis improvement

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2004-01-01

Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

Shadowfax: Moving mesh hydrodynamical integration code

Science.gov (United States)

Vandenbroucke, Bert

2016-05-01

Shadowfax simulates galaxy evolution. Written in object-oriented modular C++, it evolves a mixture of gas, subject to the laws of hydrodynamics and gravity, and any collisionless fluid only subject to gravity, such as cold dark matter or stars. For the hydrodynamical integration, it makes use of a (co-) moving Lagrangian mesh. The code has a 2D and 3D version, contains utility programs to generate initial conditions and visualize simulation snapshots, and its input/output is compatible with a number of other simulation codes, e.g. Gadget2 (ascl:0003.001) and GIZMO (ascl:1410.003).

Validation of thermohydraulic codes by comparison of experimental results with computer simulations

International Nuclear Information System (INIS)

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

1989-01-01

The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

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)

Integration of a neuroimaging processing pipeline into a pan-canadian computing grid

International Nuclear Information System (INIS)

Lavoie-Courchesne, S; Chouinard-Decorte, F; Doyon, J; Bellec, P; Rioux, P; Sherif, T; Rousseau, M-E; Das, S; Adalat, R; Evans, A C; Craddock, C; Margulies, D; Chu, C; Lyttelton, O

2012-01-01

The ethos of the neuroimaging field is quickly moving towards the open sharing of resources, including both imaging databases and processing tools. As a neuroimaging database represents a large volume of datasets and as neuroimaging processing pipelines are composed of heterogeneous, computationally intensive tools, such open sharing raises specific computational challenges. This motivates the design of novel dedicated computing infrastructures. This paper describes an interface between PSOM, a code-oriented pipeline development framework, and CBRAIN, a web-oriented platform for grid computing. This interface was used to integrate a PSOM-compliant pipeline for preprocessing of structural and functional magnetic resonance imaging into CBRAIN. We further tested the capacity of our infrastructure to handle a real large-scale project. A neuroimaging database including close to 1000 subjects was preprocessed using our interface and publicly released to help the participants of the ADHD-200 international competition. This successful experiment demonstrated that our integrated grid-computing platform is a powerful solution for high-throughput pipeline analysis in the field of neuroimaging.

Multitasking the code ARC3D. [for computational fluid dynamics

Science.gov (United States)

Barton, John T.; Hsiung, Christopher C.

1986-01-01

The CRAY multitasking system was developed in order to utilize all four processors and sharply reduce the wall clock run time. This paper describes the techniques used to modify the computational fluid dynamics code ARC3D for this run and analyzes the achieved speedup. The ARC3D code solves either the Euler or thin-layer N-S equations using an implicit approximate factorization scheme. Results indicate that multitask processing can be used to achieve wall clock speedup factors of over three times, depending on the nature of the program code being used. Multitasking appears to be particularly advantageous for large-memory problems running on multiple CPU computers.

Assessment of computer codes for VVER-440/213-type nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Szabados, L.; Ezsol, Gy.; Perneczky [Atomic Energy Research Institute, Budapest (Hungary)

1995-09-01

Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

Analysis of parallel computing performance of the code MCNP

International Nuclear Information System (INIS)

Wang Lei; Wang Kan; Yu Ganglin

2006-01-01

Parallel computing can reduce the running time of the code MCNP effectively. With the MPI message transmitting software, MCNP5 can achieve its parallel computing on PC cluster with Windows operating system. Parallel computing performance of MCNP is influenced by factors such as the type, the complexity level and the parameter configuration of the computing problem. This paper analyzes the parallel computing performance of MCNP regarding with these factors and gives measures to improve the MCNP parallel computing performance. (authors)

Linking CATHENA with other computer codes through a remote process

Energy Technology Data Exchange (ETDEWEB)

Vasic, A.; Hanna, B.N.; Waddington, G.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Girard, R. [Hydro-Quebec, Montreal, Quebec (Canada)

2005-07-01

'Full text:' CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a computer code developed by Atomic Energy of Canada Limited (AECL). The code uses a transient, one-dimensional, two-fluid representation of two-phase flow in piping networks. CATHENA is used primarily for the analysis of postulated upset conditions in CANDU reactors; however, the code has found a wider range of applications. In the past, the CATHENA thermalhydraulics code included other specialized codes, i.e. ELOCA and the Point LEPreau CONtrol system (LEPCON) as callable subroutine libraries. The combined program was compiled and linked as a separately named code. This code organizational process is not suitable for independent development, maintenance, validation and version tracking of separate computer codes. The alternative solution to provide code development independence is to link CATHENA to other computer codes through a Parallel Virtual Machine (PVM) interface process. PVM is a public domain software package, developed by Oak Ridge National Laboratory and enables a heterogeneous collection of computers connected by a network to be used as a single large parallel machine. The PVM approach has been well accepted by the global computing community and has been used successfully for solving large-scale problems in science, industry, and business. Once development of the appropriate interface for linking independent codes through PVM is completed, future versions of component codes can be developed, distributed separately and coupled as needed by the user. This paper describes the coupling of CATHENA to the ELOCA-IST and the TROLG2 codes through a PVM remote process as an illustration of possible code connections. ELOCA (Element Loss Of Cooling Analysis) is the Industry Standard Toolset (IST) code developed by AECL to simulate the thermo-mechanical response of CANDU fuel elements to transient thermalhydraulics boundary conditions. A separate ELOCA driver program

Linking CATHENA with other computer codes through a remote process

International Nuclear Information System (INIS)

Vasic, A.; Hanna, B.N.; Waddington, G.M.; Sabourin, G.; Girard, R.

2005-01-01

'Full text:' CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a computer code developed by Atomic Energy of Canada Limited (AECL). The code uses a transient, one-dimensional, two-fluid representation of two-phase flow in piping networks. CATHENA is used primarily for the analysis of postulated upset conditions in CANDU reactors; however, the code has found a wider range of applications. In the past, the CATHENA thermalhydraulics code included other specialized codes, i.e. ELOCA and the Point LEPreau CONtrol system (LEPCON) as callable subroutine libraries. The combined program was compiled and linked as a separately named code. This code organizational process is not suitable for independent development, maintenance, validation and version tracking of separate computer codes. The alternative solution to provide code development independence is to link CATHENA to other computer codes through a Parallel Virtual Machine (PVM) interface process. PVM is a public domain software package, developed by Oak Ridge National Laboratory and enables a heterogeneous collection of computers connected by a network to be used as a single large parallel machine. The PVM approach has been well accepted by the global computing community and has been used successfully for solving large-scale problems in science, industry, and business. Once development of the appropriate interface for linking independent codes through PVM is completed, future versions of component codes can be developed, distributed separately and coupled as needed by the user. This paper describes the coupling of CATHENA to the ELOCA-IST and the TROLG2 codes through a PVM remote process as an illustration of possible code connections. ELOCA (Element Loss Of Cooling Analysis) is the Industry Standard Toolset (IST) code developed by AECL to simulate the thermo-mechanical response of CANDU fuel elements to transient thermalhydraulics boundary conditions. A separate ELOCA driver program starts, ends

High-performance computational fluid dynamics: a custom-code approach

International Nuclear Information System (INIS)

Fannon, James; Náraigh, Lennon Ó; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain

2016-01-01

We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier–Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing. (paper)

High-performance computational fluid dynamics: a custom-code approach

Science.gov (United States)

Fannon, James; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain; Náraigh, Lennon Ó.

2016-07-01

We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier-Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing.

Integrating Renewable Energy Requirements Into Building Energy Codes

Energy Technology Data Exchange (ETDEWEB)

Kaufmann, John R.; Hand, James R.; Halverson, Mark A.

2011-07-01

This report evaluates how and when to best integrate renewable energy requirements into building energy codes. The basic goals were to: (1) provide a rough guide of where we’re going and how to get there; (2) identify key issues that need to be considered, including a discussion of various options with pros and cons, to help inform code deliberations; and (3) to help foster alignment among energy code-development organizations. The authors researched current approaches nationally and internationally, conducted a survey of key stakeholders to solicit input on various approaches, and evaluated the key issues related to integration of renewable energy requirements and various options to address those issues. The report concludes with recommendations and a plan to engage stakeholders. This report does not evaluate whether the use of renewable energy should be required on buildings; that question involves a political decision that is beyond the scope of this report.

New coding technique for computer generated holograms.

Science.gov (United States)

Haskell, R. E.; Culver, B. C.

1972-01-01

A coding technique is developed for recording computer generated holograms on a computer controlled CRT in which each resolution cell contains two beam spots of equal size and equal intensity. This provides a binary hologram in which only the position of the two dots is varied from cell to cell. The amplitude associated with each resolution cell is controlled by selectively diffracting unwanted light into a higher diffraction order. The recording of the holograms is fast and simple.

Users Guide to SAMINT: A Code for Nuclear Data Adjustment with SAMMY Based on Integral Experiments

Energy Technology Data Exchange (ETDEWEB)

Sobes, Vladimir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leal, Luiz C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arbanas, Goran [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-10-01

The purpose of this project is to couple differential and integral data evaluation in a continuous-energy framework. More specifically, the goal is to use the Generalized Linear Least Squares methodology employed in TSURFER to update the parameters of a resolved resonance region evaluation directly. Recognizing that the GLLS methodology in TSURFER is identical to the mathematical description of the simple Bayesian updating carried out in SAMMY, the computer code SAMINT was created to help use the mathematical machinery of SAMMY to update resolved resonance parameters based on integral data. Minimal modifications of SAMMY are required when used with SAMINT to make resonance parameter updates based on integral experimental data.

Hauser*5, a computer code to calculate nuclear cross sections

International Nuclear Information System (INIS)

Mann, F.M.

1979-07-01

HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables

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)

NADAC and MERGE: computer codes for processing neutron activation analysis data

International Nuclear Information System (INIS)

Heft, R.E.; Martin, W.E.

1977-01-01

Absolute disintegration rates of specific radioactive products induced by neutron irradition of a sample are determined by spectrometric analysis of gamma-ray emissions. Nuclide identification and quantification is carried out by a complex computer code GAMANAL (described elsewhere). The output of GAMANAL is processed by NADAC, a computer code that converts the data on observed distintegration rates to data on the elemental composition of the original sample. Computations by NADAC are on an absolute basis in that stored nuclear parameters are used rather than the difference between the observed disintegration rate and the rate obtained by concurrent irradiation of elemental standards. The NADAC code provides for the computation of complex cases including those involving interrupted irradiations, parent and daughter decay situations where the daughter may also be produced independently, nuclides with very short half-lives compared to counting interval, and those involving interference by competing neutron-induced reactions. The NADAC output consists of a printed report, which summarizes analytical results, and a card-image file, which can be used as input to another computer code MERGE. The purpose of MERGE is to combine the results of multiple analyses and produce a single final answer, based on all available information, for each element found

Poisson/Superfish codes for personal computers

International Nuclear Information System (INIS)

Humphries, S.

1992-01-01

The Poisson/Superfish codes calculate static E or B fields in two-dimensions and electromagnetic fields in resonant structures. New versions for 386/486 PCs and Macintosh computers have capabilities that exceed the mainframe versions. Notable improvements are interactive graphical post-processors, improved field calculation routines, and a new program for charged particle orbit tracking. (author). 4 refs., 1 tab., figs

FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

International Nuclear Information System (INIS)

Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

1985-12-01

FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

CAT: a computer code for the automated construction of fault trees

International Nuclear Information System (INIS)

Apostolakis, G.E.; Salem, S.L.; Wu, J.S.

1978-03-01

A computer code, CAT (Computer Automated Tree, is presented which applies decision table methods to model the behavior of components for systematic construction of fault trees. The decision tables for some commonly encountered mechanical and electrical components are developed; two nuclear subsystems, a Containment Spray Recirculation System and a Consequence Limiting Control System, are analyzed to demonstrate the applications of CAT code

Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code

International Nuclear Information System (INIS)

Duda, L.E.

1984-05-01

The Network Flow and Transport/Distributed Velocity Method (NWFT/DVM) computer code was developed primarily to fulfill a need for a computationally efficient ground-water flow and contaminant transport capability for use in risk analyses where, quite frequently, large numbers of calculations are required. It is a semi-analytic, quasi-two-dimensional network code that simulates ground-water flow and the transport of dissolved species (radionuclides) in a saturated porous medium. The development of this code was carried out under a program funded by the US Nuclear Regulatory Commission (NRC) to develop a methodology for assessing the risk from disposal of radioactive wastes in deep geologic formations (FIN: A-1192 and A-1266). In support to the methodology development program, the NRC has funded a separate Maintenance of Computer Programs Project (FIN: A-1166) to ensure that the codes developed under A-1192 or A-1266 remain consistent with current operating systems, are as error-free as possible, and have up-to-date documentations for reference by the NRC staff. Part of this effort would include verification and validation tests to assure that a code correctly performs the operations specified and/or is representing the processes or system for which it is intended. This document contains four verification problems for the NWFT/DVM computer code. Two of these problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions. The other two are code-to-code verifications where results from NWFT/DVM are compared to those of another computer code. In all cases NWFT/DVM showed good agreement with both the analytical solutions and the results from the other code

CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

International Nuclear Information System (INIS)

Krishnani, P.D.

1992-01-01

The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

International Nuclear Information System (INIS)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1990-01-01

The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

International Nuclear Information System (INIS)

2013-12-01

Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

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)

Concatenated codes for fault tolerant quantum computing

Energy Technology Data Exchange (ETDEWEB)

Knill, E.; Laflamme, R.; Zurek, W.

1995-05-01

The application of concatenated codes to fault tolerant quantum computing is discussed. We have previously shown that for quantum memories and quantum communication, a state can be transmitted with error {epsilon} provided each gate has error at most c{epsilon}. We show how this can be used with Shor`s fault tolerant operations to reduce the accuracy requirements when maintaining states not currently participating in the computation. Viewing Shor`s fault tolerant operations as a method for reducing the error of operations, we give a concatenated implementation which promises to propagate the reduction hierarchically. This has the potential of reducing the accuracy requirements in long computations.

CICART Center For Integrated Computation And Analysis Of Reconnection And Turbulence

International Nuclear Information System (INIS)

Bhattacharjee, Amitava

2016-01-01

CICART is a partnership between the University of New Hampshire (UNH) and Dartmouth College. CICART addresses two important science needs of the DoE: the basic understanding of magnetic reconnection and turbulence that strongly impacts the performance of fusion plasmas, and the development of new mathematical and computational tools that enable the modeling and control of these phenomena. The principal participants of CICART constitute an interdisciplinary group, drawn from the communities of applied mathematics, astrophysics, computational physics, fluid dynamics, and fusion physics. It is a main premise of CICART that fundamental aspects of magnetic reconnection and turbulence in fusion devices, smaller-scale laboratory experiments, and space and astrophysical plasmas can be viewed from a common perspective, and that progress in understanding in any of these interconnected fields is likely to lead to progress in others. The establishment of CICART has strongly impacted the education and research mission of a new Program in Integrated Applied Mathematics in the College of Engineering and Applied Sciences at UNH by enabling the recruitment of a tenure-track faculty member, supported equally by UNH and CICART, and the establishment of an IBM-UNH Computing Alliance. The proposed areas of research in magnetic reconnection and turbulence in astrophysical, space, and laboratory plasmas include the following topics: (A) Reconnection and secondary instabilities in large high-Lundquist-number plasmas, (B) Particle acceleration in the presence of multiple magnetic islands, (C) Gyrokinetic reconnection: comparison with fluid and particle-in-cell models, (D) Imbalanced turbulence, (E) Ion heating, and (F) Turbulence in laboratory (including fusion-relevant) experiments. These theoretical studies make active use of three high-performance computer simulation codes: (1) The Magnetic Reconnection Code, based on extended two-fluid (or Hall MHD) equations, in an Adaptive Mesh

Vectorization of nuclear codes on FACOM 230-75 APU computer

International Nuclear Information System (INIS)

Harada, Hiroo; Higuchi, Kenji; Ishiguro, Misako; Tsutsui, Tsuneo; Fujii, Minoru

1983-02-01

To provide for the future usage of supercomputer, we have investigated the vector processing efficiency of nuclear codes which are being used at JAERI. The investigation is performed by using FACOM 230-75 APU computer. The codes are CITATION (3D neutron diffusion), SAP5 (structural analysis), CASCMARL (irradiation damage simulation). FEM-BABEL (3D neutron diffusion by FEM), GMSCOPE (microscope simulation). DWBA (cross section calculation at molecular collisions). A new type of cell density calculation for particle-in-cell method is also investigated. For each code we have obtained a significant speedup which ranges from 1.8 (CASCMARL) to 7.5 (GMSCOPE), respectively. We have described in this report the running time dynamic profile analysis of the codes, numerical algorithms used, program restructuring for the vectorization, numerical experiments of the iterative process, vectorized ratios, speedup ratios on the FACOM 230-75 APU computer, and some vectorization views. (author)

Developing a friendly I/O graphical interface for the integral transport CP2D computer code

International Nuclear Information System (INIS)

Constantin, M.

2002-01-01

The code CP 2 D design and developing involved the newest methods and techniques in the first flight collision probability (FFCP) calculations. These methods are strongly connected with the computer developing both in hardware and software. The code CP 2 D was developed in INR Pitesti, between 1997-2001. It is a transport code in the first flight collision probability formalism, able to treat exactly a lot of complicated geometry (such as CANDU clusters, TRIGA and PWR fuel assemblies). The first version CP 2 D1.0 was released in 1998. The second, CP 2 D2.0, was released in 1999 and uses a multistratified coolant model (MM) for CANDU loss of coolant accident analysis. The third version, CP 2 D3.0 (2000), have incorporated a generalized burning scheme. An user-friendly graphical interface was developed in 2001. It is intended to a rapid introduction of the input data and to extract the interest information from the output files. This information is directly converted into graphics and tables contained into a single MsWord document. The introduced input data are validated by the interface if the numerical, physical and mathematical restrictions are fulfilled. The user can see the representation of the model and can interactively modify the input data until the model is correct. The interface and the code were exhaustively documented and the obtained version was released as CP 2 D4.0. The version allows to a low experienced user to build the input files, to correct the possible errors and to extract the information of interest for the analyzed problem. The paper shows the interface as a useful concept for the upgrade of the classical codes. (author)

Prodeto, a computer code for probabilistic fatigue design

Energy Technology Data Exchange (ETDEWEB)

Braam, H [ECN-Solar and Wind Energy, Petten (Netherlands); Christensen, C J; Thoegersen, M L [Risoe National Lab., Roskilde (Denmark); Ronold, K O [Det Norske Veritas, Hoevik (Norway)

1999-03-01

A computer code for structural relibility analyses of wind turbine rotor blades subjected to fatigue loading is presented. With pre-processors that can transform measured and theoretically predicted load series to load range distributions by rain-flow counting and with a family of generic distribution models for parametric representation of these distribution this computer program is available for carying through probabilistic fatigue analyses of rotor blades. (au)

Development of a graphical interface computer code for reactor fuel reloading optimization

International Nuclear Information System (INIS)

Do Quang Binh; Nguyen Phuoc Lan; Bui Xuan Huy

2007-01-01

This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)

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.

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

The failure mechanisms of HTR coated particle fuel and computer code

International Nuclear Information System (INIS)

Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

2010-01-01

The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

Status of computer codes available in AEOI for reactor physics analysis

International Nuclear Information System (INIS)

Karbassiafshar, M.

1986-01-01

Many of the nuclear computer codes available in Atomic Energy Organization of Iran AEOI can be used for physics analysis of an operating reactor or design purposes. Grasp of the various methods involved and practical experience with these codes would be the starting point for interesting design studies or analysis of operating conditions of presently existing and future reactors. A review of the objectives and flowchart of commonly practiced procedures in reactor physics analysis of LWRs and related computer codes was made, extrapolating to the nationally and internationally available resources. Finally, effective utilization of the existing facilities is discussed and called upon

Computer simulation of variform fuel assemblies using Dragon code

International Nuclear Information System (INIS)

Ju Haitao; Wu Hongchun; Yao Dong

2005-01-01

The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

COMPBRN III: a computer code for modeling compartment fires

International Nuclear Information System (INIS)

Ho, V.; Siu, N.; Apostolakis, G.; Flanagan, G.F.

1986-07-01

The computer code COMPBRN III deterministically models the behavior of compartment fires. This code is an improvement of the original COMPBRN codes. It employs a different air entrainment model and numerical scheme to estimate properties of the ceiling hot gas layer model. Moreover, COMPBRN III incorporates a number of improvements in shape factor calculations and error checking, which distinguish it from the COMPBRN II code. This report presents the ceiling hot gas layer model employed by COMPBRN III as well as several other modifications. Information necessary to run COMPBRN III, including descriptions of required input and resulting output, are also presented. Simulation of experiments and a sample problem are included to demonstrate the usage of the code. 37 figs., 46 refs

ABINIT: a computer code for matter; Abinit: un code au service de la matiere

Energy Technology Data Exchange (ETDEWEB)

Amadon, B.; Bottin, F.; Bouchet, J.; Dewaele, A.; Jollet, F.; Jomard, G.; Loubeyre, P.; Mazevet, S.; Recoules, V.; Torrent, M.; Zerah, G. [CEA Bruyeres-le-Chatel, 91 (France)

2008-07-01

The PAW (Projector Augmented Wave) method has been implemented in the ABINIT Code that computes electronic structures in atoms. This method relies on the simultaneous use of a set of auxiliary functions (in plane waves) and a sphere around each atom. This method allows the computation of systems including many atoms and gives the expression of energy, forces, stress... in terms of the auxiliary function only. We have generated atomic data for iron at very high pressure (over 200 GPa). We get a bcc-hcp transition around 10 GPa and the magnetic order disappears around 50 GPa. This method has been validated on a series of metals. The development of the PAW method has required a great effort for the massive parallelization of the ABINIT code. (A.C.)

ASTEC—the Aarhus STellar Evolution Code

Science.gov (United States)

Christensen-Dalsgaard, Jørgen

2008-08-01

The Aarhus code is the result of a long development, starting in 1974, and still ongoing. A novel feature is the integration of the computation of adiabatic oscillations for specified models as part of the code. It offers substantial flexibility in terms of microphysics and has been carefully tested for the computation of solar models. However, considerable development is still required in the treatment of nuclear reactions, diffusion and convective mixing.

CIPSS [computer-integrated process and safeguards system]: The integration of computer-integrated manufacturing and robotics with safeguards, security, and process operations

International Nuclear Information System (INIS)

Leonard, R.S.; Evans, J.C.

1987-01-01

This poster session describes the computer-integrated process and safeguards system (CIPSS). The CIPSS combines systems developed for factory automation and automated mechanical functions (robots) with varying degrees of intelligence (expert systems) to create an integrated system that would satisfy current and emerging security and safeguards requirements. Specifically, CIPSS is an extension of the automated physical security functions concepts. The CIPSS also incorporates the concepts of computer-integrated manufacturing (CIM) with integrated safeguards concepts, and draws upon the Defense Advance Research Project Agency's (DARPA's) strategic computing program

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)

Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

International Nuclear Information System (INIS)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

1983-02-01

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

Present state of the SOURCES computer code

International Nuclear Information System (INIS)

Shores, Erik F.

2002-01-01

In various stages of development for over two decades, the SOURCES computer code continues to calculate neutron production rates and spectra from four types of problems: homogeneous media, two-region interfaces, three-region interfaces and that of a monoenergetic alpha particle beam incident on a slab of target material. Graduate work at the University of Missouri - Rolla, in addition to user feedback from a tutorial course, provided the impetus for a variety of code improvements. Recently upgraded to version 4B, initial modifications to SOURCES focused on updates to the 'tape5' decay data library. Shortly thereafter, efforts focused on development of a graphical user interface for the code. This paper documents the Los Alamos SOURCES Tape1 Creator and Library Link (LASTCALL) and describes additional library modifications in more detail. Minor improvements and planned enhancements are discussed.

Comparative severe accident analysis of WWER 1000/B 320 LOCA DN100 computed by computer codes ASTEC V1.1 and SCDAP/RELAP5

International Nuclear Information System (INIS)

Kalchev, B.; Dimov, D.; Tusheva, P.; Mladenov, I.

2005-01-01

This paper presents the modelling approach for LOCA 100 mm sequence for WWER 1000-B 320 type of reactor with the integral ASTEC computer code and SCDAP/RELAP5 computer code. As a basic input deck the reference input file for Balakovo NPP from the released ASTEC CD has been applied. As a first part of the calculations for the SBLOCA sequence the ASTEC v1.1 modules CESAR, DIVA and CPA have been activated in a coupled mode. For SCDAP/RELAP5 calculation input deck for WWER 1000-B 320 has been applied which meant to be closer to the initial boundary conditions applied for ASTEC WWER 1000 input deck. A SBLOCA 100 mm comparison between ASTEC v1.1 and SCADAP/RELAP5 has been presented. ASTEC predicts vessel failure at 15620 s. ASTEC and SCDAP/RELAP5 give close but not similar results - this could be observed on the trends. The comparison of 100 mm-break shows that SCDAP/RELAP5 predicts clear phenomenological changes in primary pressure evolution and molten pool formation. Similar hydrogen production mass for both codes around 5000 s is detected

Three computer codes for safety and stability of large superconducting magnets

International Nuclear Information System (INIS)

Turner, L.R.

1985-01-01

For analyzing the safety and stability of large superconducting magnets, three computer codes TASS, SHORTURN, and SSICC have been developed, applicable to bath-cooled magnets, bath-cooled magnets with shorted turns, and magnets with internally cooled conductors respectively. The TASS code is described, and the use of the three codes is reviewed

CRACKEL: a computer code for CFR fuel management calculations

International Nuclear Information System (INIS)

Burstall, R.F.; Ball, M.A.; Thornton, D.E.J.

1975-12-01

The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)

Description of the TREBIL, CRESSEX and STREUSL computer programs, that belongs to RALLY computer code pack for the analysis of reliability systems

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1982-11-01

The RALLY computer code pack (RALLY pack) is a set of computer codes destinate to the reliability of complex systems, aiming to a risk analysis. Three of the six codes, are commented, presenting their purpose, input description, calculation methods and results obtained with each one of those computer codes. The computer codes are: TREBIL, to obtain the fault tree logical equivalent; CRESSEX, to obtain the minimal cut and the punctual values of the non-reliability and non-availability of the system; and STREUSL, for the dispersion calculation of those values around the media. In spite of the CRESSEX, in its version available at CNEN, uses a little long method to obtain the minimal cut in an HB-CNEN system, the three computer programs show good results, mainly the STREUSL, which permits the simulation of various components. (E.G.) [pt

Reducing Computational Overhead of Network Coding with Intrinsic Information Conveying

DEFF Research Database (Denmark)

Heide, Janus; Zhang, Qi; Pedersen, Morten V.

is RLNC (Random Linear Network Coding) and the goal is to reduce the amount of coding operations both at the coding and decoding node, and at the same time remove the need for dedicated signaling messages. In a traditional RLNC system, coding operation takes up significant computational resources and adds...... the coding operations must be performed in a particular way, which we introduce. Finally we evaluate the suggested system and find that the amount of coding can be significantly reduced both at nodes that recode and decode.......This paper investigated the possibility of intrinsic information conveying in network coding systems. The information is embedded into the coding vector by constructing the vector based on a set of predefined rules. This information can subsequently be retrieved by any receiver. The starting point...

Development and validation of GWHEAD, a three-dimensional groundwater head computer code

International Nuclear Information System (INIS)

Beckmeyer, R.R.; Root, R.W.; Routt, K.R.

1980-03-01

A computer code has been developed to solve the groundwater flow equation in three dimensions. The code has finite-difference approximations solved by the strongly implicit solution procedure. Input parameters to the code include hydraulic conductivity, specific storage, porosity, accretion (recharge), and initial hydralic head. These parameters may be input as varying spatially. The hydraulic conductivity may be input as isotropic or anisotropic. The boundaries either may permit flow across them or may be impermeable. The code has been used to model leaky confined groundwater conditions and spherical flow to a continuous point sink, both of which have exact analytical solutions. The results generated by the computer code compare well with those of the analytical solutions. The code was designed to be used to model groundwater flow beneath fuel reprocessing and waste storage areas at the Savannah River Plant

Multi keno-VAX a modified version of the reactor computer code Multi keno-2

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig.

Multi keno-VAX a modified version of the reactor computer code Multi keno-2

International Nuclear Information System (INIS)

Imam, M.

1995-01-01

The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig

SHEAT for PC. A computer code for probabilistic seismic hazard analysis for personal computer, user's manual

International Nuclear Information System (INIS)

Yamada, Hiroyuki; Tsutsumi, Hideaki; Ebisawa, Katsumi; Suzuki, Masahide

2002-03-01

The SHEAT code developed at Japan Atomic Energy Research Institute is for probabilistic seismic hazard analysis which is one of the tasks needed for seismic Probabilistic Safety Assessment (PSA) of a nuclear power plant. At first, SHEAT was developed as the large sized computer version. In addition, a personal computer version was provided to improve operation efficiency and generality of this code in 2001. It is possible to perform the earthquake hazard analysis, display and the print functions with the Graphical User Interface. With the SHEAT for PC code, seismic hazard which is defined as an annual exceedance frequency of occurrence of earthquake ground motions at various levels of intensity at a given site is calculated by the following two steps as is done with the large sized computer. One is the modeling of earthquake generation around a site. Future earthquake generation (locations, magnitudes and frequencies of postulated earthquake) is modeled based on the historical earthquake records, active fault data and expert judgment. Another is the calculation of probabilistic seismic hazard at the site. An earthquake ground motion is calculated for each postulated earthquake using an attenuation model taking into account its standard deviation. Then the seismic hazard at the site is calculated by summing the frequencies of ground motions by all the earthquakes. This document is the user's manual of the SHEAT for PC code. It includes: (1) Outline of the code, which include overall concept, logical process, code structure, data file used and special characteristics of code, (2) Functions of subprogram and analytical models in them, (3) Guidance of input and output data, (4) Sample run result, and (5) Operational manual. (author)

Phenomenological optical potentials and optical model computer codes

International Nuclear Information System (INIS)

Prince, A.

1980-01-01

An introduction to the Optical Model is presented. Starting with the purpose and nature of the physical problems to be analyzed, a general formulation and the various phenomenological methods of solution are discussed. This includes the calculation of observables based on assumed potentials such as local and non-local and their forms, e.g. Woods-Saxon, folded model etc. Also discussed are the various calculational methods and model codes employed to describe nuclear reactions in the spherical and deformed regions (e.g. coupled-channel analysis). An examination of the numerical solutions and minimization techniques associated with the various codes, is briefly touched upon. Several computer programs are described for carrying out the calculations. The preparation of input, (formats and options), determination of model parameters and analysis of output are described. The class is given a series of problems to carry out using the available computer. Interpretation and evaluation of the samples includes the effect of varying parameters, and comparison of calculations with the experimental data. Also included is an intercomparison of the results from the various model codes, along with their advantages and limitations. (author)

RADTRAN II: revised computer code to analyze transportation of radioactive material

International Nuclear Information System (INIS)

Taylor, J.M.; Daniel, S.L.

1982-10-01

A revised and updated version of the RADTRAN computer code is presented. This code has the capability to predict the radiological impacts associated with specific schemes of radioactive material shipments and mode specific transport variables

FIRAC: a computer code to predict fire-accident effects in nuclear facilities

International Nuclear Information System (INIS)

Bolstad, J.W.; Krause, F.R.; Tang, P.K.; Andrae, R.W.; Martin, R.A.; Gregory, W.S.

1983-01-01

FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire

Computer codes for simulating atomic-displacement cascades in solids subject to irradiation

International Nuclear Information System (INIS)

Asaoka, Takumi; Taji, Yukichi; Tsutsui, Tsuneo; Nakagawa, Masayuki; Nishida, Takahiko

1979-03-01

In order to study atomic displacement cascades originating from primary knock-on atoms in solids subject to incident radiation, the simulation code CASCADE/CLUSTER is adapted for use on FACOM/230-75 computer system. In addition, the code is modified so as to plot the defect patterns in crystalline solids. As other simulation code of the cascade process, MARLOWE is also available for use on the FACOM system. To deal with the thermal annealing of point defects produced in the cascade process, the code DAIQUIRI developed originally for body-centered cubic crystals is modified to be applicable also for face-centered cubic lattices. By combining CASCADE/CLUSTER and DAIQUIRI, we then prepared a computer code system CASCSRB to deal with heavy irradiation or saturation damage state of solids at normal temperature. Furthermore, a code system for the simulation of heavy irradiations CASCMARL is available, in which MARLOWE code is substituted for CASCADE in the CASCSRB system. (author)

Comparative analysis of the results obtained by computer code ASTEC V2 and RELAP 5.3.2 for small leak ID 80 for VVER 1000

International Nuclear Information System (INIS)

Atanasova, B.; Grudev, P.

2011-01-01

The purpose of this report is to present the results obtained by simulation and subsequent analysis of emergency mode for small leak with ID 80 for WWER 1000/B320 - Kozloduy NPP Units 5 and 6. Calculations were performed with the ASTEC v2 computer code used for calculation of severe accident, which was designed by French and German groups - IRSN and GRS. Integral RELAP5 computer code is used as a reference for comparison of results. The analyzes are focused on the processes occurring in reactor internals phase of emergency mode with significant core damage. The main thermohydraulic parameters, start of reactor core degradation and subsequent fuel relocalization till reactor vessel failure are evaluated in the analysis. RELAP5 computer code is used as a reference code to compare the results obtained till early core degradation that occurs after core stripping and excising of fuel temperature above 1200 0 C

Theoretical calculation possibilities of the computer code HAMMER

International Nuclear Information System (INIS)

Onusic Junior, J.

1978-06-01

With the aim to know the theoretical calculation possibilities of the computer code HAMMER, developed at Savanah River Laboratory, a analysis of the crytical cells assembly of the kind utilized in PWR reactors is made. (L.F.S.) [pt

INTRANS. A computer code for the non-linear structural response analysis of reactor internals under transient loads

International Nuclear Information System (INIS)

Ramani, D.T.

1977-01-01

The 'INTRANS' system is a general purpose computer code, designed to perform linear and non-linear structural stress and deflection analysis of impacting or non-impacting nuclear reactor internals components coupled with reactor vessel, shield building and external as well as internal gapped spring support system. This paper describes in general a unique computational procedure for evaluating the dynamic response of reactor internals, descretised as beam and lumped mass structural system and subjected to external transient loads such as seismic and LOCA time-history forces. The computational procedure is outlined in the INTRANS code, which computes component flexibilities of a discrete lumped mass planar model of reactor internals by idealising an assemblage of finite elements consisting of linear elastic beams with bending, torsional and shear stiffnesses interacted with external or internal linear as well as non-linear multi-gapped spring support system. The method of analysis is based on the displacement method and the code uses the fourth-order Runge-Kutta numerical integration technique as a basis for solution of dynamic equilibrium equations of motion for the system. During the computing process, the dynamic response of each lumped mass is calculated at specific instant of time using well-known step-by-step procedure. At any instant of time then, the transient dynamic motions of the system are held stationary and based on the predicted motions and internal forces of the previous instant. From which complete response at any time-step of interest may then be computed. Using this iterative process, the relationship between motions and internal forces is satisfied step by step throughout the time interval

Use of computer codes to improve nuclear power plant operation

International Nuclear Information System (INIS)

Misak, J.; Polak, V.; Filo, J.; Gatas, J.

1985-01-01

For safety and economic reasons, the scope for carrying out experiments on operational nuclear power plants (NPPs) is very limited and any changes in technical equipment and operating parameters or conditions have to be supported by theoretical calculations. In the Nuclear Power Plant Scientific Research Institute (NIIAEhS), computer codes are systematically used to analyse actual operating events, assess safety aspects of changes in equipment and operating conditions, optimize the conditions, preparation and analysis of NPP startup trials and review and amend operating instructions. In addition, calculation codes are gradually being introduced into power plant computer systems to perform real time processing of the parameters being measured. The paper describes a number of specific examples of the use of calculation codes for the thermohydraulic analysis of operating and accident conditions aimed at improving the operation of WWER-440 units at the Jaslovske Bohunice V-1 and V-2 nuclear power plants. These examples confirm that computer calculations are an effective way of solving operating problems and of further increasing the level of safety and economic efficiency of NPP operation. (author)

URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

International Nuclear Information System (INIS)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1989-01-01

The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs

A computer code for fault tree calculations: PATREC

International Nuclear Information System (INIS)

Blin, A.; Carnino, A.; Koen, B.V.; Duchemin, B.; Lanore, J.M.; Kalli, H.

1978-01-01

A computer code for evaluating the reliability of complex system by fault tree is described in this paper. It uses pattern recognition approach and programming techniques from IBM PL1 language. It can take account of many of the present day problems: multi-dependencies treatment, dispersion in the reliability data parameters, influence of common mode failures. The code is running currently since two years now in Commissariat a l'Energie Atomique Saclay center and shall be used in a future extension for automatic fault trees construction

Development of a tracer transport option for the NAPSAC fracture network computer code

International Nuclear Information System (INIS)

Herbert, A.W.

1990-06-01

The Napsac computer code predicts groundwater flow through fractured rock using a direct fracture network approach. This paper describes the development of a tracer transport algorithm for the NAPSAC code. A very efficient particle-following approach is used enabling tracer transport to be predicted through large fracture networks. The new algorithm is tested against three test examples. These demonstrations confirm the accuracy of the code for simple networks, where there is an analytical solution to the transport problem, and illustrates the use of the computer code on a more realistic problem. (author)

FRAP-T1: a computer code for the transient analysis of oxide fuel rods

International Nuclear Information System (INIS)

Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.

1977-02-01

FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer

Improvement of level-1 PSA computer code package

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

1997-07-01

This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on `The improvement of level-1 PSA Computer Codes` is divided into two main activities : (1) improvement of level-1 PSA methodology, (2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs.

Improvement of level-1 PSA computer code package

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

1997-07-01

This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The improvement of level-1 PSA Computer Codes' is divided into two main activities : 1) improvement of level-1 PSA methodology, 2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs

NEWSPEC: A computer code to unfold neutron spectra from Bonner sphere data

International Nuclear Information System (INIS)

Lemley, E.C.; West, L.

1996-01-01

A new computer code, NEWSPEC, is in development at the University of Arkansas. The NEWSPEC code allows a user to unfold, fold, rebin, display, and manipulate neutron spectra as applied to Bonner sphere measurements. The SPUNIT unfolding algorithm, a new rebinning algorithm, and the graphical capabilities of Microsoft (MS) Windows and MS Excel are utilized to perform these operations. The computer platform for NEWSPEC is a personal computer (PC) running MS Windows 3.x or Win95, while the code is written in MS Visual Basic (VB) and MS VB for Applications (VBA) under Excel. One of the most useful attributes of the NEWSPEC software is the link to Excel allowing additional manipulation of program output or creation of program input

KC-A Kinectic computer code for investigation of parametric plasma instabilities

International Nuclear Information System (INIS)

Olshansky, V.

1995-07-01

In the frame of a joint research program of the Institute of Plasma Physics of the NationaI Science Center 'Kharkov Institute of Physics and Technology' (Kh IPT), Ukraine, and the plasma physics group of the Austrian Research Center Seibersdorf (FZS) a kinetic computer code with the acronym KC for investigation of paramarametric plasma instabilities has been implemented at the computer facilities of FZS as a starting point for further research in this field. This code based on a macroparticle technique is appropriate for studying the evolution of instabilities in a turbulent plasma including saturation. The results can be of interest for heating of tokamaks of the next generation, i.g. ITER. The present report describes the underlying physical models and numerical methods as well as the code structure and how to use the code as a reference of forthcoming joint papers. (author)

Computer codes for problems of isotope and radiation research

International Nuclear Information System (INIS)

Remer, M.

1986-12-01

A survey is given of computer codes for problems in isotope and radiation research. Altogether 44 codes are described as titles with abstracts. 17 of them are in the INIS scope and are processed individually. The subjects are indicated in the chapter headings: 1) analysis of tracer experiments, 2) spectrum calculations, 3) calculations of ion and electron trajectories, 4) evaluation of gamma irradiation plants, and 5) general software

Sample test cases using the environmental computer code NECTAR

International Nuclear Information System (INIS)

Ponting, A.C.

1984-06-01

This note demonstrates a few of the many different ways in which the environmental computer code NECTAR may be used. Four sample test cases are presented and described to show how NECTAR input data are structured. Edited output is also presented to illustrate the format of the results. Two test cases demonstrate how NECTAR may be used to study radio-isotopes not explicitly included in the code. (U.K.)

Standardization of computer programs - basis of the Czechoslovak library of nuclear codes

International Nuclear Information System (INIS)

Gregor, M.

1987-01-01

A standardized form of computer code documentation has been established in the CSSR in the field of reactor safety. Structure and content of the documentation are described and codes already subject to this process are mentioned. The formation of a Czechoslovak nuclear code library and facilitated discussion of safety reports containing results of standardized codes are aimed at

Establishing a communications link between two different, incompatible, personal computers: with practical examples and illustrations and program code.

Science.gov (United States)

Davidson, R W

1985-01-01

The increasing need to communicate to exchange data can be handled by personal microcomputers. The necessity for the transference of information stored in one type of personal computer to another type of personal computer is often encountered in the process of integrating multiple sources of information stored in different and incompatible computers in Medical Research and Practice. A practical example is demonstrated with two relatively inexpensive commonly used computers, the IBM PC jr. and the Apple IIe. The basic input/output (I/O) interface chip for serial communication for each computer are joined together using a Null connector and cable to form a communications link. Using BASIC (Beginner's All-purpose Symbolic Instruction Code) Computer Language and the Disk Operating System (DOS) the communications handshaking protocol and file transfer is established between the two computers. The BASIC programming languages used are Applesoft (Apple Personal Computer) and PC BASIC (IBM Personal computer).

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

Quality assurance aspects of the computer code CODAR2

International Nuclear Information System (INIS)

Maul, P.R.

1986-03-01

The computer code CODAR2 was developed originally for use in connection with the Sizewell Public Inquiry to evaluate the radiological impact of routine discharges to the sea from the proposed PWR. It has subsequently bee used to evaluate discharges from Heysham 2. The code was frozen in September 1983, and this note gives details of its verification, validation and evaluation. Areas where either improved modelling methods or more up-to-date information relevant to CODAR2 data bases have subsequently become available are indicated; these will be incorporated in any future versions of the code. (author)

A restructuring of TF package for MIDAS computer code

International Nuclear Information System (INIS)

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

2002-01-01

TF package which defines some interpolation and extrapolation condition through user defined table has been restructured in MIDAS computer code. To do this, data transferring methods of current MELCOR code are modified and adopted into TF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of TF package addressed in this paper does module development and subroutine modification, and treats MELGEN which is making restart file as well as MELCOR which is processing calculation. The validation has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. It hints that the similar approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

FLAME: A finite element computer code for contaminant transport n variably-saturated media

International Nuclear Information System (INIS)

Baca, R.G.; Magnuson, S.O.

1992-06-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A

FLAME: A finite element computer code for contaminant transport n variably-saturated media

Energy Technology Data Exchange (ETDEWEB)

Baca, R.G.; Magnuson, S.O.

1992-06-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A.

Atmospheric dispersion of radioactive releases: Computer code DIASPORA

International Nuclear Information System (INIS)

Synodinou, B.M.; Bartzis, J.M.

1982-05-01

The computer code DIASPORA is presented. Air and ground concentrations of an airborne radioactive material released from an elevated continuous point source are calculated using Gaussian plume models. Dry and wet deposition as well as plume rise effects are taken into consideration. (author)

Analysis of the Behavior of CAREM-25 Fuel Rods Using Computer Code BACO

International Nuclear Information System (INIS)

Estevez, Esteban; Markiewicz, Mario; Marino, Armando

2000-01-01

The thermo-mechanical behavior of a fuel rod subjected to irradiation is a complex process, on which a great quantity of interrelated physical-chemical phenomena are coupled.The code BACO simulates the thermo-mechanical behavior and the evolution of fission gases of a cylindrical rod in operation.The power history of fuel rods, arising from neutronic calculations, is the program input.The code calculates, among others, the temperature distribution and the principal stresses in the pellet and cladding, changes in the porosity and restructuring of pellet, the fission gases release, evolution of the internal gas pressure.In this work some of design limits of CAREM-25's fuel rods are analyzed by means of the computer code BACO.The main variables directly related with the integrity of the fuel rod are: Maximum temperature of pellet; Cladding hoop stresses; Gases pressure in the fuel rod; Cladding axial and radial strains, etc.The analysis of results indicates that, under normal operation conditions, the maximum fuel pellet temperature, cladding stresses, pressure of gases at end of life, etc, are below the design limits considered for the fuel rod of CAREM-25 reactor

LMFBR models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

1981-10-01

Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

Benchmarking severe accident computer codes for heavy water reactor applications

Energy Technology Data Exchange (ETDEWEB)

Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

2010-07-01

Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

SALE: Safeguards Analytical Laboratory Evaluation computer code

International Nuclear Information System (INIS)

Carroll, D.J.; Bush, W.J.; Dolan, C.A.

1976-09-01

The Safeguards Analytical Laboratory Evaluation (SALE) program implements an industry-wide quality control and evaluation system aimed at identifying and reducing analytical chemical measurement errors. Samples of well-characterized materials are distributed to laboratory participants at periodic intervals for determination of uranium or plutonium concentration and isotopic distributions. The results of these determinations are statistically-evaluated, and each participant is informed of the accuracy and precision of his results in a timely manner. The SALE computer code which produces the report is designed to facilitate rapid transmission of this information in order that meaningful quality control will be provided. Various statistical techniques comprise the output of the SALE computer code. Assuming an unbalanced nested design, an analysis of variance is performed in subroutine NEST resulting in a test of significance for time and analyst effects. A trend test is performed in subroutine TREND. Microfilm plots are obtained from subroutine CUMPLT. Within-laboratory standard deviations are calculated in the main program or subroutine VAREST, and between-laboratory standard deviations are calculated in SBLV. Other statistical tests are also performed. Up to 1,500 pieces of data for each nuclear material sampled by 75 (or fewer) laboratories may be analyzed with this code. The input deck necessary to run the program is shown, and input parameters are discussed in detail. Printed output and microfilm plot output are described. Output from a typical SALE run is included as a sample problem

Regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes

International Nuclear Information System (INIS)

Vitkova, M.; Kalchev, B.; Stefanova, S.

2006-01-01

The paper presents an overview of the regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes, which are used for safety assessment of the fuel design and the fuel utilization. Some requirements to the model development, verification and validation of the codes and analysis of code uncertainties are also define. Questions concerning Quality Assurance during development and implementation of the codes as well as preparation of a detailed verification and validation plan are briefly discussed

FRANTIC: a computer code for time dependent unavailability analysis

International Nuclear Information System (INIS)

Vesely, W.E.; Goldberg, F.F.

1977-03-01

The FRANTIC computer code evaluates the time dependent and average unavailability for any general system model. The code is written in FORTRAN IV for the IBM 370 computer. Non-repairable components, monitored components, and periodically tested components are handled. One unique feature of FRANTIC is the detailed, time dependent modeling of periodic testing which includes the effects of test downtimes, test overrides, detection inefficiencies, and test-caused failures. The exponential distribution is used for the component failure times and periodic equations are developed for the testing and repair contributions. Human errors and common mode failures can be included by assigning an appropriate constant probability for the contributors. The output from FRANTIC consists of tables and plots of the system unavailability along with a breakdown of the unavailability contributions. Sensitivity studies can be simply performed and a wide range of tables and plots can be obtained for reporting purposes. The FRANTIC code represents a first step in the development of an approach that can be of direct value in future system evaluations. Modifications resulting from use of the code, along with the development of reliability data based on operating reactor experience, can be expected to provide increased confidence in its use and potential application to the licensing process

Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

International Nuclear Information System (INIS)

Wren, D.J.; Popov, N.; Snell, V.G.

2004-01-01

Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

Computer code to simulate transients in a steam generator of PWR nuclear power plants

International Nuclear Information System (INIS)

Silva, J.M. da.

1979-01-01

A digital computer code KIBE was developed to simulate the transient behavior of a Steam Generator used in Pressurized Water Reactor Power PLants. The equations of Conservation of mass, energy and momentum were numerically integrated by an implicit method progressively in the several axial sections into which the Steam Generator was divided. Forced convection heat transfer was assumed on the primary side, while on the secondary side all the different modes of heat transfer were permitted and deternined from the various correlations. The stability of the stationary state was verified by its reproducibility during the integration of the conservation equation without any pertubation. Transient behavior resulting from pertubations in the flow and the internal energy (temperature) at the inlet of the primary side were simulated. The results obtained exhibited satisfactory behaviour. (author) [pt

LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

International Nuclear Information System (INIS)

Bordner, G.L.

1979-10-01

The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

SIMCRI: a simple computer code for calculating nuclear criticality parameters

International Nuclear Information System (INIS)

Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.

1986-03-01

This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)

Analysis of the Length of Braille Texts in English Braille American Edition, the Nemeth Code, and Computer Braille Code versus the Unified English Braille Code

Science.gov (United States)

Knowlton, Marie; Wetzel, Robin

2006-01-01

This study compared the length of text in English Braille American Edition, the Nemeth code, and the computer braille code with the Unified English Braille Code (UEBC)--also known as Unified English Braille (UEB). The findings indicate that differences in the length of text are dependent on the type of material that is transcribed and the grade…

A study to compute integrated dpa for neutron and ion irradiation environments using SRIM-2013

Science.gov (United States)

Saha, Uttiyoarnab; Devan, K.; Ganesan, S.

2018-05-01

Displacements per atom (dpa), estimated based on the standard Norgett-Robinson-Torrens (NRT) model, is used for assessing radiation damage effects in fast reactor materials. A computer code CRaD has been indigenously developed towards establishing the infrastructure to perform improved radiation damage studies in Indian fast reactors. We propose a method for computing multigroup neutron NRT dpa cross sections based on SRIM-2013 simulations. In this method, for each neutron group, the recoil or primary knock-on atom (PKA) spectrum and its average energy are first estimated with CRaD code from ENDF/B-VII.1. This average PKA energy forms the input for SRIM simulation, wherein the recoil atom is taken as the incoming ion on the target. The NRT-dpa cross section of iron computed with "Quick" Kinchin-Pease (K-P) option of SRIM-2013 is found to agree within 10% with the standard NRT-dpa values, if damage energy from SRIM simulation is used. SRIM-2013 NRT-dpa cross sections applied to estimate the integrated dpa for Fe, Cr and Ni are in good agreement with established computer codes and data. A similar study carried out for polyatomic material, SiC, shows encouraging results. In this case, it is observed that the NRT approach with average lattice displacement energy of 25 eV coupled with the damage energies from the K-P option of SRIM-2013 gives reliable displacement cross sections and integrated dpa for various reactor spectra. The source term of neutron damage can be equivalently determined in the units of dpa by simulating self-ion bombardment. This shows that the information of primary recoils obtained from CRaD can be reliably applied to estimate the integrated dpa and damage assessment studies in accelerator-based self-ion irradiation experiments of structural materials. This study would help to advance the investigation of possible correlations between the damages induced by ions and reactor neutrons.

Accuracy assessment of a new Monte Carlo based burnup computer code

International Nuclear Information System (INIS)

El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.

2012-01-01

Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.

(Nearly) portable PIC code for parallel computers

International Nuclear Information System (INIS)

Decyk, V.K.

1993-01-01

As part of the Numerical Tokamak Project, the author has developed a (nearly) portable, one dimensional version of the GCPIC algorithm for particle-in-cell codes on parallel computers. This algorithm uses a spatial domain decomposition for the fields, and passes particles from one domain to another as the particles move spatially. With only minor changes, the code has been run in parallel on the Intel Delta, the Cray C-90, the IBM ES/9000 and a cluster of workstations. After a line by line translation into cmfortran, the code was also run on the CM-200. Impressive speeds have been achieved, both on the Intel Delta and the Cray C-90, around 30 nanoseconds per particle per time step. In addition, the author was able to isolate the data management modules, so that the physics modules were not changed much from their sequential version, and the data management modules can be used as open-quotes black boxes.close quotes

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

CICART Center For Integrated Computation And Analysis Of Reconnection And Turbulence

Energy Technology Data Exchange (ETDEWEB)

Bhattacharjee, Amitava [Univ. of New Hampshire, Durham, NH (United States)

2016-03-27

CICART is a partnership between the University of New Hampshire (UNH) and Dartmouth College. CICART addresses two important science needs of the DoE: the basic understanding of magnetic reconnection and turbulence that strongly impacts the performance of fusion plasmas, and the development of new mathematical and computational tools that enable the modeling and control of these phenomena. The principal participants of CICART constitute an interdisciplinary group, drawn from the communities of applied mathematics, astrophysics, computational physics, fluid dynamics, and fusion physics. It is a main premise of CICART that fundamental aspects of magnetic reconnection and turbulence in fusion devices, smaller-scale laboratory experiments, and space and astrophysical plasmas can be viewed from a common perspective, and that progress in understanding in any of these interconnected fields is likely to lead to progress in others. The establishment of CICART has strongly impacted the education and research mission of a new Program in Integrated Applied Mathematics in the College of Engineering and Applied Sciences at UNH by enabling the recruitment of a tenure-track faculty member, supported equally by UNH and CICART, and the establishment of an IBM-UNH Computing Alliance. The proposed areas of research in magnetic reconnection and turbulence in astrophysical, space, and laboratory plasmas include the following topics: (A) Reconnection and secondary instabilities in large high-Lundquist-number plasmas, (B) Particle acceleration in the presence of multiple magnetic islands, (C) Gyrokinetic reconnection: comparison with fluid and particle-in-cell models, (D) Imbalanced turbulence, (E) Ion heating, and (F) Turbulence in laboratory (including fusion-relevant) experiments. These theoretical studies make active use of three high-performance computer simulation codes: (1) The Magnetic Reconnection Code, based on extended two-fluid (or Hall MHD) equations, in an Adaptive Mesh

Integration of the TNXYZ computer program inside the platform Salome

International Nuclear Information System (INIS)

Chaparro V, F. J.

2014-01-01

The present work shows the procedure carried out to integrate the code TNXYZ as a calculation tool at the graphical simulation platform Salome. The TNXYZ code propose a numerical solution of the neutron transport equation, in several groups of energy, steady-state and three-dimensional geometry. In order to discretized the variables of the transport equation, the code uses the method of discrete ordinates for the angular variable, and a nodal method for the spatial dependence. The Salome platform is a graphical environment designed for building, editing and simulating mechanical models mainly focused on the industry and unlike other software, in order to form a complete scheme of pre and post processing of information, to integrate and control an external source code. Before the integration the in the Salome platform TNXYZ code was upgraded. TNXYZ was programmed in the 90s using Fortran 77 compiler; for this reason the code was adapted to the characteristics of the current Fortran compilers; in addition, with the intention of extracting partial results over the process sequence, the original structure of the program underwent a modularization process, i.e. the main program was divided into sections where the code performs major operations. This procedure is controlled by the information module (YACS) on Salome platform, and it could be useful for a subsequent coupling with thermal-hydraulics codes. Finally, with the help of the Monte Carlo code Serpent several study cases were defined in order to check the process of integration; the verification process consisted in performing a comparison of the results obtained with the code executed as stand-alone and after modernized, integrated and controlled by the Salome platform. (Author)

Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

International Nuclear Information System (INIS)

2014-02-01

The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

Nuclear model codes available at the Nuclear Energy Agency Computer Program Library (NEA-CPL)

International Nuclear Information System (INIS)

Sartori, E.; Garcia Viedma, L. de

1976-01-01

This paper briefly outlines the objectives of the NEA-CPL and its activities in the field of Nuclear Model Computer Codes. A short description of the computer codes available from the CPL in this field is also presented. (author)

Application of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the OECD/NRC BWR turbine trip benchmark and its performance on multi-processor computers

International Nuclear Information System (INIS)

Langenbuch, S.; Schmidt, K.D.; Velkov, K.

2003-01-01

The OECD/NRC BWR Turbine Trip (TT) Benchmark is investigated to perform code-to-code comparison of coupled codes including a comparison to measured data which are available from turbine trip experiments at Peach Bottom 2. This Benchmark problem for a BWR over-pressure transient represents a challenging application of coupled codes which integrate 3-dimensional neutron kinetics into thermal-hydraulic system codes for best-estimate simulation of plant transients. This transient represents a typical application of coupled codes which are usually performed on powerful workstations using a single CPU. Nowadays, the availability of multi-CPUs is much easier. Indeed, powerful workstations already provide 4 to 8 CPU, computer centers give access to multi-processor systems with numbers of CPUs in the order of 16 up to several 100. Therefore, the performance of the coupled code Athlet-Quabox/Cubbox on multi-processor systems is studied. Different cases of application lead to changing requirements of the code efficiency, because the amount of computer time spent in different parts of the code is varying. This paper presents main results of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the BWR TT Benchmark together with evaluations of the code performance on multi-processor computers. (authors)

Selection of a computer code for Hanford low-level waste engineered-system performance assessment

International Nuclear Information System (INIS)

McGrail, B.P.; Mahoney, L.A.

1995-10-01

Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites

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

PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code

International Nuclear Information System (INIS)

Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.

1976-12-01

Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems

Study and application of Dot 3.5 computer code in radiation shielding problems

International Nuclear Information System (INIS)

Otto, A.C.; Mendonca, A.G.; Maiorino, J.R.

1983-01-01

The application of nuclear transportation code S sub(N), Dot 3.5, to radiation shielding problems is revised. Aiming to study the better available option (convergence scheme, calculation mode), of DOT 3.5 computer code to be applied in radiation shielding problems, a standard model from 'Argonne Code Center' was selected and a combination of several calculation options to evaluate the accuracy of the results and the computational time was used, for then to select the more efficient option. To illustrate the versatility and efficacy in the application of the code for tipical shielding problems, the streaming neutrons calculation along a sodium coolant channel is ilustrated. (E.G.) [pt

A k-distribution-based radiation code and its computational optimization for an atmospheric general circulation model

International Nuclear Information System (INIS)

Sekiguchi, Miho; Nakajima, Teruyuki

2008-01-01

The gas absorption process scheme in the broadband radiative transfer code 'mstrn8', which is used to calculate atmospheric radiative transfer efficiently in a general circulation model, is improved. Three major improvements are made. The first is an update of the database of line absorption parameters and the continuum absorption model. The second is a change to the definition of the selection rule for gas absorption used to choose which absorption bands to include. The last is an upgrade of the optimization method used to decrease the number of quadrature points used for numerical integration in the correlated k-distribution approach, thereby realizing higher computational efficiency without losing accuracy. The new radiation package termed 'mstrnX' computes radiation fluxes and heating rates with errors less than 0.6 W/m 2 and 0.3 K/day, respectively, through the troposphere and the lower stratosphere for any standard AFGL atmospheres. A serious cold bias problem of an atmospheric general circulation model using the ancestor code 'mstrn8' is almost solved by the upgrade to 'mstrnX'

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

TBCI and URMEL - New computer codes for wake field and cavity mode calculations

International Nuclear Information System (INIS)

Weiland, T.

1983-01-01

Wake force computation is important for any study of instabilities in high current accelerators and storage rings. These forces are generated by intense bunches of charged particles passing cylindrically symmetric structures on or off axis. The adequate method for computing such forces is the time domain approach. The computer Code TBCI computes for relativistic as well as for nonrelativistic bunches of arbitrary shape longitudinal and transverse wake forces up to the octupole component. TBCI is not limited to cavity-like objects and thus applicable to bellows, beam pipes with varying cross sections and any other nonresonant structures. For the accelerating cavities one also needs to know the resonant modes and frequencies for the study of instabilities and mode couplers. The complementary code named URMEL computes these fields for any azimuthal dependence of the fields in ascending order. The mathematical procedure being used is very safe and does not miss modes. Both codes together represent a unique tool for accelerator design and are easy to use

Independent verification and validation testing of the FLASH computer code, Versiion 3.0

International Nuclear Information System (INIS)

Martian, P.; Chung, J.N.

1992-06-01

Independent testing of the FLASH computer code, Version 3.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at various Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Verification tests, and validation tests, were used to determine the operational status of the FLASH computer code. These tests were specifically designed to test: correctness of the FORTRAN coding, computational accuracy, and suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: blind testing, independent applications, and graduated difficulty of test cases. Both quantitative and qualitative testing was performed through evaluating relative root mean square values and graphical comparisons of the numerical, analytical, and experimental data. Four verification test were used to check the computational accuracy and correctness of the FORTRAN coding, and three validation tests were used to check the suitability to simulating actual conditions. These tests cases ranged in complexity from simple 1-D saturated flow to 2-D variably saturated problems. The verification tests showed excellent quantitative agreement between the FLASH results and analytical solutions. The validation tests showed good qualitative agreement with the experimental data. Based on the results of this testing, it was concluded that the FLASH code is a versatile and powerful two-dimensional analysis tool for fluid flow. In conclusion, all aspects of the code that were tested, except for the unit gradient bottom boundary condition, were found to be fully operational and ready for use in hydrological and environmental studies

Integration of rocket turbine design and analysis through computer graphics

Science.gov (United States)

Hsu, Wayne; Boynton, Jim

1988-01-01

An interactive approach with engineering computer graphics is used to integrate the design and analysis processes of a rocket engine turbine into a progressive and iterative design procedure. The processes are interconnected through pre- and postprocessors. The graphics are used to generate the blade profiles, their stacking, finite element generation, and analysis presentation through color graphics. Steps of the design process discussed include pitch-line design, axisymmetric hub-to-tip meridional design, and quasi-three-dimensional analysis. The viscous two- and three-dimensional analysis codes are executed after acceptable designs are achieved and estimates of initial losses are confirmed.

Latest improvements on TRACPWR six-equations thermohydraulic code

International Nuclear Information System (INIS)

Rivero, N.; Batuecas, T.; Martinez, R.; Munoz, J.; Lenhardt, G.; Serrano, P.

1999-01-01

The paper presents the latest improvements on TRACPWR aimed at adapting the code to present trends on computer platforms, architectures and training requirements as well as extending the scope of the code itself and its applicability to other technologies different from Westinghouse PWR one. Firstly major features of TRACPWR as best estimate and real time simulation code are summed, then the areas where TRACPWR is being improved are presented. These areas comprising: (1) Architecture: integrating TRACPWR and RELAP5 codes, (2) Code scope enhancement: modelling the Mid-Loop operation, (3) Code speed-up: applying parallelization techniques, (4) Code platform downswing: porting to Windows N1 platform, (5) On-line performance: allowing simulation initialisation from a Plant Process Computer, and (6) Code scope extension: using the code for modelling VVER and PHWR technology. (author)

Neutron spectrum unfolding using computer code SAIPS

International Nuclear Information System (INIS)

Karim, S.

1999-01-01

The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results. (author)

Compilation of documented computer codes applicable to environmental assessment of radioactivity releases

International Nuclear Information System (INIS)

Hoffman, F.O.; Miller, C.W.; Shaeffer, D.L.; Garten, C.T. Jr.; Shor, R.W.; Ensminger, J.T.

1977-04-01

The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title

Computer code for the atomistic simulation of lattice defects and dynamics

International Nuclear Information System (INIS)

Schiffgens, J.O.; Graves, N.J.; Oster, C.A.

1980-04-01

The computer code COMENT used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect properties, defect migration, and defect stability. This report documents Version IV of COMENT (models, methods, and implementation) and defines current code options. Version IV of COMENT generates only face-centered-cubic (fcc) crystal lattices. However, an effort was made to structure COMENT to allow addition of new options with a minimum of change in the existing version of the code. This document describes the calling program and thirty-two user-defined subroutines. Fourteen subroutines (ALOYORD, DASPKA, DFCT, DSLOAN, DSLOIN, EXPAND, POT1, POT2, POT3, POT4, POT5, POT6, POT7, and THRMAL) are associated with the selection of program options; only a few of these are used in any given analysis. Seven of the other subroutines (CRYSTL, IEAF, INCBOX, LABLE, MINILAT, SPEFORS, and SQUEZ are used to establish a variety of arrays and conditions required for each analysis; most of them are used once in a given calculation. The remaining eleven subroutines (DAMP, DIRECT, IDDEF, NEAF, INBIN, FILBIN, FTBIN, PAC3, UNPAC3, PACF, and UNPACF) are used many times in each calculation; the last eight of these are used many times in each time step during the integration and, therefore, are written in COMPASS (CDC assembly language). The COMPASS subroutines are described in sufficient detail to permit easy conversion to some other assembly language or to FORTRAN

Computer code to assess accidental pollutant releases

International Nuclear Information System (INIS)

Pendergast, M.M.; Huang, J.C.

1980-07-01

A computer code was developed to calculate the cumulative frequency distributions of relative concentrations of an air pollutant following an accidental release from a stack or from a building penetration such as a vent. The calculations of relative concentration are based on the Gaussian plume equations. The meteorological data used for the calculation are in the form of joint frequency distributions of wind and atmospheric stability

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

Validating the BISON fuel performance code to integral LWR experiments

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

2016-05-15

Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

Computer codes to assess risks from nuclear power plants with LWR's

International Nuclear Information System (INIS)

Alonso, A.; Blanco, J.; Francia, L.; Gallego, E.; Morales, L.; Ortega, P.; Torres, C.

1986-01-01

The codes used to quantify risks from nuclear power plants are described. For QRA level 1 (quantitative risk assessment) qualitative and quantitative codes are described. Codes to estimate uncertainties, importance and dependent failures are also included. For QRA-level 2, the most important codes dealing with thermohydraulics, molten core and aerosols behaviour are described. For QRA-level 3 the list includes integrated as well as separate models. Only light water reactors are considered. The presentation is general but the authors describe with more detail those codes they are more familiar with or the ones they have created through their research effort. (author)

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

MQRAD, a computer code for synchrotron radiation from quadrupole magnets

International Nuclear Information System (INIS)

Morimoto, Teruhisa.

1984-01-01

The computer code, MQRAD, is developed for the calculation of the synchrotron radiation from the particles passing through quadrupole magnets at the straight section of the electron-positron colliding machine. This code computes the distributions of photon numbers and photon energies at any given points on the beam orbit. In this code, elements such as the quadrupole magnets and the drift spaces can be divided into many sub-elements in order to obtain the results with good accuracy. The synchrotron radiation produced by inserted quadrupole magnets at the interaction region of the electron-positron collider is one of the main background sources to the detector. The masking system against the synchrotron radiation at TRISTAN is very important because of the relatively high beam energy and the long straight section, which are 30 GeV and 100 meters, respectively. MQRAD has been used to design the masking system of the TOPAZ detector and the result is presented here as an example. (author)

Proceedings of the conference on computer codes and the linear accelerator community

International Nuclear Information System (INIS)

Cooper, R.K.

1990-07-01

The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned

Proceedings of the conference on computer codes and the linear accelerator community

Energy Technology Data Exchange (ETDEWEB)

Cooper, R.K. (comp.)

1990-07-01

The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned.

Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

International Nuclear Information System (INIS)

Raymond, P.; Caruge, D.; Paik, H.J.

1994-01-01

The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

User Instructions for the Systems Assessment Capability, Rev. 0, Computer Codes Volume 1: Inventory, Release, and Transport Modules

International Nuclear Information System (INIS)

Eslinger, Paul W.; Engel, David W.; Gerhardstein, Lawrence H.; Lopresti, Charles A.; Nichols, William E.; Strenge, Dennis L.

2001-12-01

One activity of the Department of Energy's Groundwater/Vadose Zone Integration Project is an assessment of cumulative impacts from Hanford Site wastes on the subsurface environment and the Columbia River. Through the application of a system assessment capability (SAC), decisions for each cleanup and disposal action will be able to take into account the composite effect of other cleanup and disposal actions. The SAC has developed a suite of computer programs to simulate the migration of contaminants (analytes) present on the Hanford Site and to assess the potential impacts of the analytes, including dose to humans, socio-cultural impacts, economic impacts, and ecological impacts. The general approach to handling uncertainty in the SAC computer codes is a Monte Carlo approach. Conceptually, one generates a value for every stochastic parameter in the code (the entire sequence of modules from inventory through transport and impacts) and then executes the simulation, obtaining an output value, or result. This document provides user instructions for the SAC codes that handle inventory tracking, release of contaminants to the environment, and transport of contaminants through the unsaturated zone, saturated zone, and the Columbia River

Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center. [Radiation transport codes

Energy Technology Data Exchange (ETDEWEB)

McGill, B.; Maskewitz, B.F.; Anthony, C.M.; Comolander, H.E.; Hendrickson, H.R.

1976-01-01

The term ''code package'' is used to describe a miscellaneous grouping of materials which, when interpreted in connection with a digital computer, enables the scientist--user to solve technical problems in the area for which the material was designed. In general, a ''code package'' consists of written material--reports, instructions, flow charts, listings of data, and other useful material and IBM card decks (or, more often, a reel of magnetic tape) on which the source decks, sample problem input (including libraries of data) and the BCD/EBCDIC output listing from the sample problem are written. In addition to the main code, and any available auxiliary routines are also included. The abstract format was chosen to give to a potential code user several criteria for deciding whether or not he wishes to request the code package. (RWR)

Fuel rod computations. The COMETHE code in its CEA version

International Nuclear Information System (INIS)

Lenepveu, Dominique.

1976-01-01

The COMETHE code (COde d'evolution MEcanique et THermique) is intended for computing the irradiation behavior of water reactor fuel pins. It is concerned with steadily operated cylindrical pins, containing fuel pellet stacks (UO 2 or PuO 2 ). The pin consists in five different axial zones: two expansion chambers, two blankets, and a central core that may be divided into several stacks parted by plugs. As far as computation is concerned, the pin is divided into slices (maximum 15) in turn divided into rings (maximum 50). Information are obtained for each slice: the radial temperature distribution, heat transfer coefficients, thermal flux at the pin surface, changes in geometry according to temperature conditions, and specific burn-up. The physical models involved take account for: heat transfer, fission gas release, fuel expansion, and creep of the can. Results computed with COMETHE are compared with those from ELP and EPEL irradiation experiments [fr

Integrated State Estimation and Contingency Analysis Software Implementation using High Performance Computing Techniques

Energy Technology Data Exchange (ETDEWEB)

Chen, Yousu; Glaesemann, Kurt R.; Rice, Mark J.; Huang, Zhenyu

2015-12-31

Power system simulation tools are traditionally developed in sequential mode and codes are optimized for single core computing only. However, the increasing complexity in the power grid models requires more intensive computation. The traditional simulation tools will soon not be able to meet the grid operation requirements. Therefore, power system simulation tools need to evolve accordingly to provide faster and better results for grid operations. This paper presents an integrated state estimation and contingency analysis software implementation using high performance computing techniques. The software is able to solve large size state estimation problems within one second and achieve a near-linear speedup of 9,800 with 10,000 cores for contingency analysis application. The performance evaluation is presented to show its effectiveness.

EXTRAN: A computer code for estimating concentrations of toxic substances at control room air intakes

International Nuclear Information System (INIS)

Ramsdell, J.V.

1991-03-01

This report presents the NRC staff with a tool for assessing the potential effects of accidental releases of radioactive materials and toxic substances on habitability of nuclear facility control rooms. The tool is a computer code that estimates concentrations at nuclear facility control room air intakes given information about the release and the environmental conditions. The name of the computer code is EXTRAN. EXTRAN combines procedures for estimating the amount of airborne material, a Gaussian puff dispersion model, and the most recent algorithms for estimating diffusion coefficients in building wakes. It is a modular computer code, written in FORTRAN-77, that runs on personal computers. It uses a math coprocessor, if present, but does not require one. Code output may be directed to a printer or disk files. 25 refs., 8 figs., 4 tabs

Integration of adaptive process control with computational simulation for spin-forming

International Nuclear Information System (INIS)

Raboin, P. J. LLNL

1998-01-01

Improvements in spin-forming capabilities through upgrades to a metrology and machine control system and advances in numerical simulation techniques were studied in a two year project funded by Laboratory Directed Research and Development (LDRD) at Lawrence Livermore National Laboratory. Numerical analyses were benchmarked with spin-forming experiments and computational speeds increased sufficiently to now permit actual part forming simulations. Extensive modeling activities examined the simulation speeds and capabilities of several metal forming computer codes for modeling flat plate and cylindrical spin-forming geometries. Shape memory research created the first numerical model to describe this highly unusual deformation behavior in Uranium alloys. A spin-forming metrology assessment led to sensor and data acquisition improvements that will facilitate future process accuracy enhancements, such as a metrology frame. Finally, software improvements (SmartCAM) to the manufacturing process numerically integrate the part models to the spin-forming process and to computational simulations

Monocrystal sputtering by the computer simulation code ACOCT

International Nuclear Information System (INIS)

Yamamura, Yasunori; Takeuchi, Wataru.

1987-09-01

A new computer code ACOCT has been developed in order to simulate the atomic collisions in the crystalline target within the binary collision approximation. The present code is more convenient as compared with the MARLOWE code, and takes the higher-order simultaneous collisions into account. To cheke the validity of the ACOCT program, we have calculated sputtering yields for various ion-target combinations and compared with the MARLOWE results. It is found that the calculated yields by the ACOCT program are in good agreements with those by the MARLOWE code. The ejection patterns of sputtered atoms were also calculated for the major surfaces of fcc, bcc, diamond and hcp structures, and we have got reasonable agreements with experimental results. In order to know the effects of the simultaneous collision in the slowing down process the sputtering yields and the projected ranges are calculated, changeing the parameter of the criterion for the simultaneous collision, and the effect of the simultaneous collision is found to depend on the crystal orientation. (author)

Mother code specifications (Appendix to CEA report 2472)

International Nuclear Information System (INIS)

Pillard, Denise; Soule, Jean-Louis

1964-12-01

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

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

International Nuclear Information System (INIS)

Prado-Herrero, P.

1992-01-01

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

COMTA - a computer code for fuel mechanical and thermal analysis

International Nuclear Information System (INIS)

Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.

1979-01-01

COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)

AUTOET code (a code for automatically constructing event trees and displaying subsystem interdependencies)

International Nuclear Information System (INIS)

Wilson, J.R.; Burdick, G.R.

1977-06-01

This is a user's manual for AUTOET I and II. AUTOET I is a computer code for automatic event tree construction. It is designed to incorporate and display subsystem interdependencies and common or key component dependencies in the event tree format. The code is written in FORTRAN IV for the CDC Cyber 76 using the Integrated Graphics System (IGS). AUTOET II incorporates consequence and risk calculations, in addition to some other refinements. 5 figures

Integrating bar-code devices with computerized MC and A systems

International Nuclear Information System (INIS)

Anderson, L.K.; Boor, M.G.; Hurford, J.M.

1998-01-01

Over the past seven years, Los Alamos National Laboratory developed several generations of computerized nuclear materials control and accountability (MC and A) systems for tracking and reporting the storage, movement, and management of nuclear materials at domestic and international facilities. During the same period, Oak Ridge National Laboratory was involved with automated data acquisition (ADA) equipment, including installation of numerous bar-code scanning stations at various facilities to serve as input devices to computerized systems. Bar-code readers, as well as other ADA devices, reduce input errors, provide faster input, and allow the capture of data in remote areas where workstations do not exist. Los Alamos National Laboratory and Oak Ridge National Laboratory teamed together to implement the integration of bar-code hardware technology with computerized MC and A systems. With the expertise of both sites, the two technologies were successfully merged with little difficulty. Bar-code input is now available with several functions of the MC and A systems: material movements within material balance areas (MBAs), material movements between MBAs, and physical inventory verification. This paper describes the various components required for the integration of these MC and A systems with the installed bar-code reader devices and the future directions for these technologies

A molecular dynamics simulation code ISIS

International Nuclear Information System (INIS)

Kambayashi, Shaw

1992-06-01

Computer simulation based on the molecular dynamics (MD) method has become an important tool complementary to experiments and theoretical calculations in a wide range of scientific fields such as physics, chemistry, biology, and so on. In the MD method, the Newtonian equations-of-motion of classical particles are integrated numerically to reproduce a phase-space trajectory of the system. In the 1980's, several new techniques have been developed for simulation at constant-temperature and/or constant-pressure in convenient to compare result of computer simulation with experimental results. We first summarize the MD method for both microcanonical and canonical simulations. Then, we present and overview of a newly developed ISIS (Isokinetic Simulation of Soft-spheres) code and its performance on various computers including vector processors. The ISIS code has a capability to make a MD simulation under constant-temperature condition by using the isokinetic constraint method. The equations-of-motion is integrated by a very accurate fifth-order finite differential algorithm. The bookkeeping method is also utilized to reduce the computational time. Furthermore, the ISIS code is well adopted for vector processing: Speedup ratio ranged from 16 to 24 times is obtained on a VP2600/10 vector processor. (author)

LWR-WIMS, a computer code for light water reactor lattice calculations

International Nuclear Information System (INIS)

Halsall, M.J.

1982-06-01

LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

SWAAM-LT: The long-term, sodium/water reaction analysis method computer code

International Nuclear Information System (INIS)

Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.

1993-01-01

The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-05-01

This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

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.

A code to compute borehole fluid conductivity profiles with multiple feed points

International Nuclear Information System (INIS)

Hale, F.V.; Tsang, C.F.

1988-03-01

It is of much current interest to determine the flow characteristics of fractures intersecting a wellbore in order to understand the hydrologic behavior of fractured rocks. Often inflow from these fractures into the wellbore is at very low rates. A new procedure has been proposed and a corresponding method of analysis developed to obtain fracture inflow parameters from a time sequence of electric conductivity logs of the borehole fluid. The present report is a companion document to NTB--88-13 giving the details of equations and computer code used to compute borehole fluid conductivity distributions. Verification of the code used and a listing of the code are also given. (author) 9 refs., 5 figs., 7 tabs

The MESORAD dose assessment model: Computer code

International Nuclear Information System (INIS)

Ramsdell, J.V.; Athey, G.F.; Bander, T.J.; Scherpelz, R.I.

1988-10-01

MESORAD is a dose equivalent model for emergency response applications that is designed to be run on minicomputers. It has been developed by the Pacific Northwest Laboratory for use as part of the Intermediate Dose Assessment System in the US Nuclear Regulatory Commission Operations Center in Washington, DC, and the Emergency Management System in the US Department of Energy Unified Dose Assessment Center in Richland, Washington. This volume describes the MESORAD computer code and contains a listing of the code. The technical basis for MESORAD is described in the first volume of this report (Scherpelz et al. 1986). A third volume of the documentation planned. That volume will contain utility programs and input and output files that can be used to check the implementation of MESORAD. 18 figs., 4 tabs

Protect Heterogeneous Environment Distributed Computing from Malicious Code Assignment

Directory of Open Access Journals (Sweden)

V. S. Gorbatov

2011-09-01

Full Text Available The paper describes the practical implementation of the protection system of heterogeneous environment distributed computing from malicious code for the assignment. A choice of technologies, development of data structures, performance evaluation of the implemented system security are conducted.

Acceleration of a Monte Carlo radiation transport code

International Nuclear Information System (INIS)

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

1996-01-01

Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics

ENDF/B Pre-Processing Codes: Implementing and testing on a Personal Computer

International Nuclear Information System (INIS)

McLaughlin, P.K.

1987-05-01

This document describes the contents of the diskettes containing the ENDF/B Pre-Processing codes by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a series of 7 diskettes. (author)

Computer code for calculating personnel doses due to tritium exposures

International Nuclear Information System (INIS)

Graham, C.L.; Parlagreco, J.R.

1977-01-01

This report describes a computer code written in LLL modified Fortran IV that can be used on a CDC 7600 for calculating personnel doses due to internal exposures to tritium. The code is capable of handling various exposure situations and is also capable of detecting a large variety of data input errors that would lead to errors in the dose assessment. The critical organ is the body water

RADTRAN: a computer code to analyze transportation of radioactive material

International Nuclear Information System (INIS)

Taylor, J.M.; Daniel, S.L.

1977-04-01

A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry

MELCOR computer code manuals: Primer and user's guides, Version 1.8.3 September 1994. Volume 1

International Nuclear Information System (INIS)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users' Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

Recommendations for computer code selection of a flow and transport code to be used in undisturbed vadose zone calculations for TWRS immobilized wastes environmental analyses

International Nuclear Information System (INIS)

VOOGD, J.A.

1999-01-01

An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis

Pre-Service Teachers' Perception of Quick Response (QR) Code Integration in Classroom Activities