International Nuclear Information System (INIS)

A combination of computer codes such as LATREP, HWRAXAV and CITATION is utilized in an attempt to analyze the nuclear design characteristics of the CAXDU-PHWR of the Wolsung Unit 1. The major nuclear properties to be computed are the lattice properties of CANDU fuel channel and the core channel power distribution. The computed results are compared with the preliminary safety reports documentation for the Wolsung reactor. The observed discrepancies between our computations and the preliminary safety reports values are discussed in terms of incomplete information on the description of the core configuration in the preliminary safety reports and the different calculation methods. (author).

Science.gov (United States)

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer ...

Energy Technology Data Exchange (ETDEWEB)

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW ...

International Nuclear Information System (INIS)

An overview of LOCA analysis in Japanese industry is presented. The BASH-M code, developed for large scale LOCA reflooding analysis, is given as an example of verification and improvement of US computer programs are given. The code's application to the operational safety analysis concerns the following main areas: 1D drift flux model base computer program CANAC; CANAC-based advanced training simulator; emergency operating procedures. The author considers also the code application to the following new PWR safety design concepts: use of steam generators for decay heat removal at LOCA conditions; use of horizontal type steam generator for maintaining two-phase natural circulation under the reactor coolant system submerged. 9 figs.

Energy Technology Data Exchange (ETDEWEB)

In the thermal hydraulics codes developed for fire safety analysis and for containment thermal hydraulic analysis, junctions in the multi-compartment geometries is often modeled as uni-directional junctions. However, ceiling junctions are known to depict unstable/oscillatory bi-directional flow behavior. Detailed investigations have been carried out to understand the unstable flow behaviour of a junction by analyzing an earlier reported experiment and its subsequent two dimensional numerical RANS based study of fire in an enclosure. The authors attempt more realistic and desired three dimensional and inherently transient large eddy simulations using a computer code Fire Dynamics Simulator (FDS). The paper presents the details of the analysis, the results obtained and further studies required to be conducted so that the findings can be applied to the fire/containment thermal hydraulics analysis ...

International Nuclear Information System (INIS)

More than twenty years after commencing commercial operation in 1983, Wolsong Unit 1(W1- NPP), the first CANDU Pressurized Heavy Water Reactor (PHWR) in Korea, has been undergoing refurbishment. Safety analyses were required to evaluate the safety of W1-NPP because significant amount of equipment has been refurbished. To evaluate the effectiveness of W1-NPP after these upgrades, new safety analyses were performed using the same technical standards of Wolsong Units 2, 3, 4 (W234-NPP) for Design Basis Accidents (DBA). The refurbished W1- NPP is expected to be licensed for full power operation based on the verified safety analysis results that are obtained by using the upgraded computer codes and newly adopted technical standards of W234-NPP

Energy Technology Data Exchange (ETDEWEB)

For light water reactors, loss of coolant is an important point in safety analysis, whereas for gas-cooled reactors the ingress of water into the core region is an incident of safety relevance. The applicability of the computer code system GAMTEREX to pebble beds of spherical high-temperature gas-cooled reactor fuel elements with simulated water ingress is verified by experiment. The measurements were performed at a Siemens-Argonaut reactor, using its ring core as a driver zone for a pebble-bed core in the center of the reactor.

International Nuclear Information System (INIS)

The in-core reload design procedure, computer codes, design contents and comparison between expected design values and measured values for Daya Bay Nuclear Power Station are presented. The in-core reload design for Daya Bay Nuclear Power Station performed by Nuclear Power Institute of China satisfies the requirements of safety and economy. The expected design values are fit well with measured values

Energy Technology Data Exchange (ETDEWEB)

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW ...

Energy Technology Data Exchange (ETDEWEB)

With the recent advance of the utilization of nuclear energy in a large scale, social interest is being focussed in the potential risk which the nuclear technology will accompany. Especially after the accidents in Chernobyl and other nuclear facilities, serious anxiety to the utilization of nuclear energy is prevailing among the general public. In order to meet the anxiety and distrust of the population in the use of the nuclear power, the health effect or risk which radioactive materials released into the environment will bring about should be comprehensively and properly evaluated, and then should be widely reported to the population. The development of HESANS code system (Health and Environmental Safety Assessment Network System) was planned to set up such a comprehensive computer code that covers a whole pathway of radioactive material from its release to estimates of derived health effects in the ...

International Nuclear Information System (INIS)

This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, ...

International Nuclear Information System (INIS)

SMART (System integrated Modular Advanced ReacTor) , is under development at the Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection system (SIS), and an adoption of 4 trains of passive residual heat removal system (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a steam generator tube rupture (SGTR) is one of the most important initiating events which results in a high core damage frequency. Clear understanding of accident progression with various combinations of the safety systems helps ...

Energy Technology Data Exchange (ETDEWEB)

Methods are discussed for optimization of organizational model of longwall mining. Operations at a working face, their duration, number of miners, miner qualifications are analyzed. The so-called method for allocation of heteregeneous resources is used. Using the method optimum organizational models of working faces are constructed. The models consider working conditions, coal output, personnel, miner qualifications, duration of various operations and of idletime, and safety regulations. A computer code developed by the author for optimizing organizational models of longwall mining is described. 9 refs.

British Library Electronic Table of Contents (United Kingdom)

Piping in nuclear power plants is vital to the proper operation and safety of these facilities. To assure safety in the unlikely event of a pipe break, it is necessary to evaluate the consequences from the resulting whipping pipe on neighboring components and structures. Numerical simulations allow for rapid evaluation of these consequences. Before simulations can be accepted, however, the methodology and computer codes must be validated against experimental results. This paper uses a probabilistic approach to validate pipe whip simulations against limited experimental results. Probabilistic analysis software was developed and coupled to existing deterministic finite element software. An example of a whipping pipe impacting against a reinforced concrete slab was simulated. The described pr...

International Nuclear Information System (INIS)

The paper presents the procedure of an investigation of WWER-1000 primary circuit temperature-stress field by the use of thermohydraulic computation data for a pressurized thermal shock event ''Core overcooling''. The procedure is based on a model of the plane stress state with ideal contact between wall and medium for the calculation. The computation data are calculated on the base of WWER-1000 thermohydraulic model by the RELAP5/MOD3 codes. This model was developed jointly by the Bulgarian and BNL/USA staff to provide an analytical tool for performing safety analysis. As a result of calculations by codes the computation data for temperature field law (linear laws of a few distinguished parts) and pressure of coolant at points on inner surface of WWER-1000 primary circuit equipment are received. Such calculations can be used as a base for determination of ...

Energy Technology Data Exchange (ETDEWEB)

The Next Generation Nuclear Plant (NGNP), a demonstration reactor and hydrogen production facility proposed for construction at the INEEL, is expected to be a high-temperature gas-cooled reactor (HTGR). Computer codes used in design and safety analysis for the NGNP must be benchmarked against experimental data. The INEEL and ANL have examined information about several past and present experimental and prototypical facilities based on HTGR concepts to assess the potential of these facilities for use in this benchmarking effort. Both reactors and critical facilities applicable to pebble-bed and prismatic block-type cores have been considered. Four facilities--HTR-PROTEUS, HTR-10, ASTRA, and AVR--appear to have the greatest potential for use in benchmarking codes for pebble-bed reactors. Similarly, for the prismatic block-type reactor design, two experiments have been ranked as having the highest ...

Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: Hydrogen release and distribution in nuclear power plant containment is an important safety issue. Selection of a proper turbulence model is important for accurate estimation of the mixing process. The selection of turbulence model is dictated by the best compromise between accuracy and computational efforts. For this, three different turbulence models, viz. Standard k-{epsilon}, RNG k-{epsilon} and Reynolds Stress Model, based on Reynolds averaged Navier Stokes equations (RANS) approach, were used. The computations were done using the CFD code FLUENT, which is based on the control volume methodology. The computational results were compared with the experimental results of HYMIS test facility, where helium was used to simulate hydrogen. The processes of helium plume rise, multiple plume merging, distribution and mixing were studied. Based on these ...

International Nuclear Information System (INIS)

Uranium Hexafluoride is a material used in the various processes which comprise the front end of the nuclear fuel cycle (conversion, enrichment and fuel fabrication). Confinement of UF6 is a very important safety requirement since this material is highly reactive and presents safety hazards to humans. The present paper discusses the safety relevant aspects of accidental releases of UF6 inside process confinement buildings. Postulated accidental scenarios are analyzed and their consequences evaluated. Implant releases rates are estimated using computer code predictions. A time dependent homogeneous compartment model is used to predict concentrations of UF6, hydrogen fluoride and uranyl fluoride inside a confinement building, as well as to evaluate source terms released to the atmosphere. These source terms can be used as input to atmospheric dispersion models to evaluate consequences ...

International Nuclear Information System (INIS)

A computer code based on Nodal expansion method has been developed for solving two groups three dimensional diffusion equation. This code can be used for fuel management and time average core calculation. Explicit Xenon and fuel temperature estimation are also incorporated in this code. TAPP-4 phase-B physics experimental results were analyzed using this code and a code based on FD method. This paper gives the comparison of the observed data and the results obtained with this code and FD code. (author)

Science.gov (United States)

In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a ...

Energy Technology Data Exchange (ETDEWEB)

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

International Nuclear Information System (INIS)

Designs of nuclear reactor plants aim for high performance under safety consideration. Because of large scale and high pressure/temperature conditions, data from costly mockup tests have been required to verify simulation codes of systems and components. Establishment of design by analysis (DBA) in nuclear engineering is required for development of next generation nuclear reactors. Recent powerful computers and simulation technique enable numerical analyses to predict realistic behaviors of thermo-fluid flow, structure and do on. The present report describes resent simulation results of complex gas-liquid two-phase flow, large scale structure dynamics and fluid-structure interaction. (author)

British Library Electronic Table of Contents (United Kingdom)

Steel building frames are often analyzed for stability in an elastic way, while most of their columns behave inelastically at the buckling stage. Most column design provisions allow for inelastic behavior, but overall inelastic stability analysis is rarely performed. In this study the analysis philosophy is centered on the overall frame stability and its true safety factor. As many columns show inelastic behavior at the buckling stage, the proposed procedure takes due consideration of this fact. Once the overall buckling factor for the frame is obtained, individual column effective length factors, and their true slenderness ratios are computed, and used in the design relationships. This procedure circumvents the use of design nomographs and numerous formulas proposed in the past to allevia...

Energy Technology Data Exchange (ETDEWEB)

Various physics applications involve the computation of the standard hypergeometric function {sub 2}F{sub 1} and its derivatives. Because it is not an intrinsic in the common programming languages, automatic differentiation tools will either differentiate through the code that computes {sub 2}F{sub 1} if that code is available or require the user to provide a hand-written derivative code. We present options for the derivative computation in the context of an ionization problem and compare the approach implemented in the Diamant library to standard methods.

International Nuclear Information System (INIS)

This paper will review code and standard and the safety related features of major components of Monju: Components of the Reactor Coolant Boundary; Components of the Reactor Shurdown Systems; Components of the Decay Heat Removal Systems; Components of the Engineered Safety Features; Other Safety Related Components. Their relationship to the system or plant function is emphasized, in reviewing these components.

International Nuclear Information System (INIS)

The FFTF is a 400-MWt sodium-cooled fast neutron flux test reactor located on the US government-owned Hanford Site in southeastern Washington state. The reactor is operated for the US Department of Energy by the Westinghouse Hanford Company. Since FFTF started routine operation in 1982, the commercially fabricated driver fuel has performed flawlessly to well beyond the design goal peak burnup of 80 MWd/kgM. The core average discharge exposure is now some 60% beyond the original design expectations and attests to the ruggedness and reliability of the mixed oxide fuel system. In Cycle 9 sixteen long-life assemblies were installed to begin the irradiation of mixed oxides in the advanced low-swelling alloy HT-9 as the Core Demonstration Experiment (CDE). Operation of the plant from initial startup testing to ten cycles of operation has confirmed that the nuclear characteristics are well within the design predictions, and all parameters have remained inside the operating envelope defined by ...

Energy Technology Data Exchange (ETDEWEB)

To resolve the central thermal safety issue for spent fuel dry storage concrete canister design or Wolsung (CANDU) nuclear power plant unit 1, a thermal analysis method has been developed for the complicated geometry of rod bundles and the multi-dimensional and multi-mode heat transfer phenomena. The canister geometry is simplified and combined heat transfer by conduction, convection, and radiation is considered through effective heat transfer coefficients. Mean temperature distributions of the fuel bundles within the fuel basket are obtained by solving the heat transfer problem using an existing computer code HEATING5. The measured steady state temperature distribution within a mock-up of a storage basket is compared to the calculated result. Steady state and/or transient fuel temperature distributions have been calculated for various ambient conditions at the canister exterior surface.

International Nuclear Information System (INIS)

Four very common long term mineralogical changes in salt formations are discussed in the view of the safety considerations for underground repositories. Two of these processes, the 'Hartsalz' and 'Carnallite' dissolution were studied in two scale in situ experiments. The results are presented and compared with the results of the geochemical modelling with the computer code EQ3/6. Furthermore the reactions leading to the formation of the gypsum cap rock on the top of the Zechstein salt formations and to the polyhalitization of anhydrite are discussed. Geological field observations and mineral assemblages agree well with the results of the geochemical modelling employing the Pitzer formalism along with the Harvie, Moller and Weare database. We conclude that once the mechanisms of the chemical reactions are well understood it becomes possible to evaluate realistically whether such processes, when encountered in the repository, ...

Energy Technology Data Exchange (ETDEWEB)

The Stockpile Stewardship Program (SSP) is a single, highly integrated technical program for maintaining the safety and reliability of the U.S. nuclear stockpile. The SSP uses past nuclear test data along with current and future nonnuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production capability to support stockpile requirements. This integrated national program requires the continued use of current facilities and programs along with new experimental facilities and computational enhancements to support these programs. The Advanced Simulation and Computing Program (ASC)1 is a cornerstone of the SSP, providing simulation capabilities and computational resources to support the ...

British Library Electronic Table of Contents (United Kingdom)

To estimate the success criteria of an operators action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the `ASME PRA standard' also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called ...

International Nuclear Information System (INIS)

Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactor and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core ...

Energy Technology Data Exchange (ETDEWEB)

The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of ...

British Library Electronic Table of Contents (United Kingdom)

Entorhinal grid cells in mammals fire as a function of animal location, with spatially periodic response patterns. This nonlocal periodic representation of location, a local variable, is unlike other neural codes. There is no theoretical explanation for why such a code should exist. We examined how accurately the grid code with noisy neurons allows an ideal observer to estimate location and found this code to be a previously unknown type of population code with unprecedented robustness to noise. In particular, the representational accuracy attained by grid cells over the coding range was in a qualitatively different class from what is possible with observed sensory and motor population codes. We found that a simple neural network can effectively correct the grid code. To the best of our kn...

International Nuclear Information System (INIS)

A nuclear power plant engineering (NPP) simulator may include many system functions such as thermal-hydraulic calculation, 3D reactor neutron kinetics model, control and protection system, display and operational human-machine interface, intelligent alarm system, etc. If all those functions are simulated using a single code, the size of the program will cause structural or managerial problems, and the hardware requirement could be tremendous. Thus, any local function errors or changes may affect the whole code, which will make development and maintenance extremely costly. So the nondistributed code is not considered to be flexible and feasible. Currently, distributed and parallel technologies have been applied to develop NPPs engineering simulator for safety analysis, verification of advanced main control room display, operation and intelligent alarm design. The distributed simulation divides a huge ...

Energy Technology Data Exchange (ETDEWEB)

FLUTAN is a highly vectorized computer code for 3-D fluiddynamic and thermal-hydraulic analyses in cartesian and cylinder coordinates. It is related to the family of COMMIX codes originally developed at Argonne National Laboratory, USA. To a large extent, FLUTAN relies on basic concepts and structures imported from COMMIX-1B and COMMIX-2 which were made available to KfK in the frame of cooperation contracts in the fast reactor safety field. While on the one hand not all features of the original COMMIX versions have been implemented in FLUTAN, the code on the other hand includes some essential innovative options like CRESOR solution algorithm, general 3-dimensional rebalacing scheme for solving the pressure equation, and LECUSSO-QUICK-FRAM techniques suitable for reducing `numerical diffusion` in both the enthalphy and momentum equations. This report provides users with detailed ...

International Nuclear Information System (INIS)

The safety analysis checklist scope and the criteria of key parameters needed to be checked for Daya bay reload cycles are introduced. The INCORE code package was used for the safety evaluation of Daya bay unit 2 cycle 2. The method and the contents can not only be applicable for Daya Bay reload cycles but also for Qinshan 600 MW and Qinshan 300 MW reload cycles.

Energy Technology Data Exchange (ETDEWEB)

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

International Nuclear Information System (INIS)

This paper describes the benchmark analysis results of Rapid Boron Dilution (RBD) transient tests. The RBD transient tests were carried out at University of Maryland. The data were obtained as a part of the International Standard Problem No. 43, ISP-43, for code assessment. Nuclear Power Engineering Corporation (NUPEC) participated the benchmark analysis of ISP443 with the PLASHY code. The PLASHY code is a general purpose single-phase fluid analysis code, developed by NUPEC. The code has two types of module, Cartesian/Cylindrical coordinate system module and BFC module. In order to reduce the computing time, the Cartesian/Cylindrical coordinate system module was employed. All calculations were done on a parallel computer, IBM SP2, by using the 12 of 76 CPU units. (author)

Energy Technology Data Exchange (ETDEWEB)

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of ...

Science.gov (United States)

The Heavy Water Reactor Facility (HWRF), as part of the defense-in-depth philosophy to mitigate the effect of design-basis and severe accidents, is equipped with a passive containment cooling system (PCCS). The function of the PCCS is to provide a safety-grade path to the ultimate heat sink for the removal of the reactor coolant system sensible heat and core decay heat. Ambient air enters an annular space between the steel containment shell and the surrounding concrete shield building through inlets in the shield building wall, is heated via natural convection, rises, and exits the building through a chimney located above the containment dome. A test program is in place to access parameters important to the effective operation of the PCCS. This paper focuses on the large-scale tests (LSTs). The objectives of these tests are as follows: (1) demonstrate natural circulation cooling with more prototypic cylinder and dome surface area ratios than were available in the ...

Energy Technology Data Exchange (ETDEWEB)

The O({alpha}) electroweak radiative corrections to {gamma}{gamma}{yields}WW{yields}4f within the electroweak standard model are calculated in double-pole approximation (DPA). Virtual corrections are treated in DPA, leading to a classification into factorizable and non-factorizable contributions, and real-photonic corrections are based on complete lowest-order matrix elements for {gamma}{gamma}{yields}4f+{gamma}. Soft and collinear singularities appearing in the virtual and real corrections are combined alternatively in two different ways, namely by using the dipole subtraction method or by applying phase-space slicing. The radiative corrections are implemented in a Monte Carlo generator called Coffer {gamma}{gamma} - the computer code can be obtained from the authors upon request - which optionally includes anomalous triple and quartic gauge-boson couplings in addition and performs a convolution over realistic spectra of the photon beams. A ...

International Nuclear Information System (INIS)

Hydrogen combustion phenomenology during severe accidents in nuclear power plants has been a safety issue. Especially, flame acceleration (FA) and Deflagration-to-Detonation Transition (DDT) are important because of their possible destructive impact on the containment or plant systems in it. Accordingly, it is the design goal to avoid FA and DDT for the hydrogen mitigation system. As a result from extensive effort dedicated to resolve the hydrogen issue, compiled information on the FA and DDT has been provided by the NEA. The FA criterion was suggested in terms of the mixture expansion ratio ? , and the criterion for onset of DDT was based on the greatness of the geometrical size of the reactive system compared with the detonation cell width ? of the average mixture composition. The DDT onset criterion reflects the current state of knowledge based on the experimental database; however, the criterion is not sufficient but a necessary condition. Therefore, it should ...

Energy Technology Data Exchange (ETDEWEB)

Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain ...

Energy Technology Data Exchange (ETDEWEB)

PRESTO-II is a computer code developed to evaluate possible health effects from shallow land disposal of low level radioactive wastes. PRESTO-PREP is a data preprocessor that has been developed to expedite the formation of input data sets for PRESTO-II. PRESTO-PREP utilizes a library of nuclide and risk-specific data. Given an initial waste inventory, the code creates the radionuclide portion of the associated input data set for PRESTO-II. 2 references.

Energy Technology Data Exchange (ETDEWEB)

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.) 4 refs.

Energy Technology Data Exchange (ETDEWEB)

Subcritical experiments using californium source-driven noise analysis (CSDNA) and Feynman variance-to-mean methods were performed with an alpha-phase plutonium sphere reflected by nickel shells, up to a maximum thickness of 7.62 cm. Both methods provide means of determining the subcritical multiplication of a system containing nuclear material. A benchmark analysis of the experiments was performed for inclusion in the 2010 edition of the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Benchmark models have been developed that represent these subcritical experiments. An analysis of the computed eigenvalues and the uncertainty in the experiment and methods was performed. The eigenvalues computed using the CSDNA method were very close to those calculated using MCNP5; however, computed eigenvalues are used in the analysis of the CSDNA method. Independent calculations using ...

Energy Technology Data Exchange (ETDEWEB)

This report covers application of Argonne National Laboratory`s (ANL`s) computer codes to simulation and analysis of components of the magnetohydrodynamic (MHD) power train system at the Component Development and Integration Facility (CDIF). Major components of the system include a 50-MWt coal-fired, two-stage combustor and an MHD channel. The combustor, designed and built by TRW, includes a deswirl section between the first and the second-stage combustor and a converging nozzle following the second-stage combustor, which connects to the MHD channel. ANL used computer codes to simulate and analyze flow characteristics in various components of the MHD system. The first-stage swirl combustor was deemed a mature technology and, therefore, was not included in the computer simulation. Several versions of the ICOMFLO computer code were used for ...

International Nuclear Information System (INIS)

The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref.

International Nuclear Information System (INIS)

FLOWER, a computer code recommended by USNRC for assessing the environmental impact in tidal regions, was adapted and improved so as to be suitable to deal with the influence of drift stream along seashore to the dilution of contaminants and heat in the bay mouth. And the code outputs were presented with more mid-results such as average concentrations and temperature values for all tides considered. Finally, the modified code is applied to the dispersion calculation of heat and liquid effluents from Daya Bay Nuclear Power Plant, and the impacts from routine operation of the plant on Daya Bay sea waters were given.

Energy Technology Data Exchange (ETDEWEB)

The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.).

Energy Technology Data Exchange (ETDEWEB)

On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory ...

Energy Technology Data Exchange (ETDEWEB)

A three-dimensional (3D) thermal-hydraulic model and a numerical procedure for the simulation and analysis of a steady-state, as well as transient operation of nuclear power plant components are presented. A two-fluid approach is applied to modelling of two-phase flow. Thermal-hydraulics of a horizontal steam generator in the WWER 1000 nuclear power plant has been simulated at the full load, steady-state operation. A comparison of the numerical results with data measured at the NPP Novovoronjezh shows good agreement. 3D numerical results can be used in plant design or retrofitting, in nuclear power plant operation and safety analysis and as improvement of existing one-dimensional thermal-hydraulics models of the horizontal steam generator which are assessed by system codes used for the nuclear power plant safety analyses. (author)

International Nuclear Information System (INIS)

The use of computer codes based on Monte Carlo methods to perform criticality calculations has become common-place. Although results frequently published in the literature report calculated k_e_f_f values to four decimal places, people who use the codes in their everyday work say that they only believe the first two decimal places of any result. The lack of confidence in the computed k_e_f_f values may be due to the tendency of the reported standard deviation to underestimate errors associated with the Monte Carlo process. The standard deviation as reported by the codes is the standard deviation of the mean of the k_e_f_f values for individual generations in the computer simulation, not the standard deviation of the computed k_e_f_f value compared with the physical system. A more subtle problem with the standard deviation of the mean as ...

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WIPS (Whip and Impact of Piping Systems) is a special purpose computer code for the structural analysis of pipe whip dynamic effects following a postulated pipe rupture. WIPS has been developed primarily to provide support for the pipe whip analysis procedures described in Section 3.6.2 of the US Nuclear Regulatory Commission Standard Review Plan. This report summarizes the purpose and scope of the WIPS development effort, identifying those clauses in the Standard Review Plan which refer to pipe whip analysis, and indicating how the WIPS code can be used to provide supporting data. Detailed information on use of the code is contained in accompanying reports which cover: (1) user instructions; (2) theory; (3) programming procedures; and (4) verification examples.

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The ASFIT (Anisotropic Source-Flux Iteration Technique) code was first developed in India about 1970 to solve the radiation transport equation represented in the form of coupled integral equations separating the spatial and energy-angular transmissions. The ASFIT code uses a nodal structure in the wavelength [lambda] domain in Compton units (CU) rather than in the energy domain. The ASFIT-VARI code is the latest version available from the Radiation Shielding Information Center. It incorporates variable dimensioning and has been adapted for use on MS-DOS personal computers using the Ryan-McFarland or Microsoft Version 5.0 FORTRAN compilers. While earlier versions used point cross sections (well suited for gamma-ray transport), the present version also allows multigroup cross sections for neutron and coupled neutron-gamma-ray transport.

International Nuclear Information System (INIS)

Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76"o C which is much less than the limiting maximum value of fuel temperature of ...

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The purpose of this paper is to describe the embedded computer systems approach taken at Experimental Breeder Reactor II (EBR-II) for non-safety related systems. The hardware and software structures for typical embedded systems are presented The embedded systems development process is described. Three examples are given which illustrate typical embedded computer applications in EBR-II.

Science.gov (United States)

FIRESET is a PC-based computer code which calculates current as a function of time for an RLC circuit containing up to fifteen series conductors which undergo rapid heating and subsequent explosion as a consequence of an electric current which passes through them. In its original form, the code was developed to model electrical waveforms measured when a large, typically 25.4 x 25.4 x 0.051-mm, aluminum foil was exploded using a capacitor bank with tens of kilojoules of stored energy. The code proved to be useful for this purpose, and it was recognized that it could also be used for modeling the electrical response of detonator bridgewires. In view of the increasing use of slapper detonators for DOD applications, we wish to make the latest version of the code, available to DOD laboratories and contractors for use in designing firing systems which employ slapper or exploding ...

International Nuclear Information System (INIS)

The IAEA Code of Practice on Quality Assurance for Safety in Nuclear Power Plants and the corresponding Safety Guides are reviewed and compared with quality assurance (QA) practices in the IAEA Member States. The QA requirements stipulated by the Code place on the nuclear power plant owner the responsibility to establish an overall QA programme for the plant. In selecting the QA programme level for specific activities, the Code allows of a flexible approach but does not specify gradation in programme requirements. The Code is placing the burden of quality-achieving and quality-assuring functions on the task-performing organizations, namely the designers, manufacturers, constructors and plant operators. The plant owner provides for the management of the overall QA programme, surveillance of activities and verifications of the effectiveness of the constituent ...

Energy Technology Data Exchange (ETDEWEB)

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the 'ASME PRA standard' also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS ...

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The report provides and overview of calculation models for the simulation of airflows and deals comprehensively with field and multi-zone models as well as the coupling of individual zone and multi-zone models. Examples of calculations are given. figs., tabs., refs.

Science.gov (United States)

include modeling geometric complexities, coupling between the flow and the .... noted that the AGARD 445.6 wing data set, which has been widely used for .... higher-level computational aeroelastic code validations at transonic flow separation ...

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FLOWTRAN-TF is a new two-component, two-phase thermal-hydraulics code to capture the detailed assembly behavior associated with loss-of-coolant accident analyses in multichannel assemblies of the SRS reactors. The local interfacial area of the two-phase mixture is computed by summing the interfacial areas contributed by each of three flow regimes. For smooth flow regime transitions, the code uses an interpolation technique in terms of component void fraction for each basic flow regime.

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Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system built February, 1992, by W.S. Parks. These modules are used for burnup, cooling, separate, and edit calculations.

International Nuclear Information System (INIS)

The method of syndrome coding for data compression read out from multiwire proportional chambers that has been previously proposed is generalized in case of its application to registration of the coordinates of events detected. The questions of execution of arithmetic and algebraic operations on the Galois field elements and their hardware implementation are considered. The method of computation is presented of a specialized processor for parallel computing the coordinates of three sparks. The estimate of its speed is equal to 185 ns. Data compression, data selection and coordinate calculations are performed without use of memory elements and timing pulses.

Energy Technology Data Exchange (ETDEWEB)

This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P{sub N}) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the ...

International Nuclear Information System (INIS)

A safety analysis code, named as SPACE, for a pressurized water reactor is under development to obtain a licensing to be used for the PWR design and to hold entire proprietary rights. The task of KAERI is to develop the physical models and correlations which are required to solve the field equations. It can be divided into four parts; i) flow regime determination, ii) wall heat transfer, iii) wall and interfacial friction, iv) interfacial heat and mass transfer. This paper will describe the process to develop the models for the two-phase flow regime maps in the horizontal and vertical pipes.

International Nuclear Information System (INIS)

Wolsong 2,3 and 4 fuel channel analysis during a large break loss of coolant accident with loss of ECCS injection (LOCA/LOECC) is performed to obtain the heat load to moderator. Because the single channel analysis requires the establishment of the safety codes and their input decks, the present study follows the same safety analysis methodology found in FSAR of Wolsong 2,3 and 4. From this work we obtain the safety tools such as CATHENA MOD3.5b/Rev.1 and CHAN-II/A MOD2 codes, and their code modeling in a form of code input deck. The analysis consists of two parts: front-end (blowdown period) and back-end. For the front-end analysis the fuel and pressure tube (PT) temperatures, and PT circumferential strains at the end of front-end as well as fuel channel depressurization are calculated using CATHENA code and used as ...

International Nuclear Information System (INIS)

On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB, the U.S. agency responsible for determining the cause of transportation accidents), to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory ...

Science.gov (United States)

This paper presents results of solid-liquid phase change, driven by volumetric energy generation, in a vertical cylinder. We show excellent agreement between a quasi-static, approximate analytical solution valid for Stefan numbers less than one, and a computational model solved using the CFD code FLUENT®. A computational study also shows the effect that the volumetric energy generation has on both the mushy zone thickness and convection in the melt during phase change.

Science.gov (United States)

The accelerating flow of a lighter continuous phase through a heavier one is considered. Small nonuniformities grow into large ones due to the Rayleigh-Taylor instability. An experiment exemplifying the large bubble formation due to Rayleigh-Taylor instability was performed and simulated using the PHOENICS 84 computer code. The same numerical procedure was applied to the two-phase flow in a gun barrel. It shows that the acceleration provided by the movement of the projectile can cause initial nonuniformities to grow with time.

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This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not ...

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Boron Neutron Capture Therapy (BNCT) is a promising binary treatment modality for high-grade primary brain tumors (glioblastoma multiforme, GM) and other cancers. BNCT employs a boron-10 containing compound that preferentially accumulates in the cancer cells in the brain. Upon neutron capture by {sup 10}B energetic alpha particles and triton released at the absorption site kill the cancer cell. In order to gain penetration depth in the brain Fairchild proposed, for this purpose, the use of energetic epithermal neutrons at about 10 keV. Phase I/II clinical trials of BNCT for GM are underway at the Brookhaven Medical Research Reactor (BMRR) and at the MIT Reactor, using these nuclear reactors as the source for epithermal neutrons. In light of the limitations of new reactor installations, e.g. cost, safety and licensing, and limited capability for modulating the reactor based neutron beam energy spectra alternative neutron sources are being contemplated for wider ...

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code to design a pressurized water reactor and to obtain the licences including entire proprietary rights is under development in domestic research and development project. The purpose and scope of this report is to develop the flow regimes related models for inter-phase friction, wall frictions, wall heat transfer, and inter-phase heat and mass transfer in two-phase three-field equations. In order to choose choose the flow regime criteria, we have investigated various exiting best-estimate T/H codes in this chapter 2. They are the RELAP5-3D, TRAC-M, CATHARE, MARS codes. Around 500 references used in these codes have been collected and reviewed. Also we have investigated eleven papers in detail. In chapter 3, based on the selected flow regimes, the flow regime maps for a gas-liquid flow in horizontal and vertical tubes have decided including the mechanisms of flow ...

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A stable, accurate, and efficient implementation of MacCormack's explicit algorithm for the Parabolized Navier-Stokes equations is demonstrated. The familiar problem of decoding the conservative axial flux vector is solved, resulting in accurate, smooth dependent variable profiles through the viscous-layer sonic line. Source terms due to transformation of the parabolized governing equations into the computational plane and the equations into the computational plane and the resulting metric differencing have been identified and eliminated through inclusion of appropriate geometric conservation law terms. Test cases computed include two- and three-dimensional supersonic and hypersonic flow at laminar and turbulent Reynolds numbers. The computed results demonstrate very good agreement with experiment and with solutions of the full Navier-Stokes equations. Computational times ...

Science.gov (United States)

Prepared to make housing project managers, maintenance engineering staff, and designers and architects of HUD - assisted and HUD - insured housing projects and mobile home parks aware of their responsibilities under the Natural Gas Pipeline Safety Act, this handbook provides technical guidance for the reduction of gas leaks and the handling of gas leak incidents in residential areas. The hazards of natural gas, concern over gas safety in residential areas, applicable codes and standards, and sources of information and help are reviewed, along with maintenance and engineering considerations relating to gas pipe failures, the detection of gas leaks, appliance servicing, and pipe installation, repair, and replacement. Housing management considerations are also examined, with attention to recordkeeping and reporting, emergency contingency planning, tenant education, staff qualifications and training, and the development and use ...

Wastenet

... My road safety has come a long way since a bus driver yelled 'Go and read the Highway Code!' at me in my first wobbly week. I've got over my bike theft paranoia and happily leave my trusty steed shackled to lampposts around London. And I've figured out the optimum skirt-length for pedaling - neither flasher short nor so long it gets stuck in the oily spokes. But ...

Science.gov (United States)

This thesis study covers both global performance and local phenomena analyses focusing on natural circulation reactor design safety. Four important topics are included: the global SBWR design safety assessment, important local phenomena investigation, steady and transient natural circulation process study, and two-phase instability analysis. The conceptual design of the SBWR-200 is introduced in this thesis and the global performance of a natural circulation reactor is then assessed using PUMA integral test data and RELAP5 simulations. A safety assessment methodology is developed to evaluate the PUMA integral test data extrapolation and code scalability. The RELAP5 code simulation capability in low-pressure low-flow conditions is also validated. The study shows that the code is capable of predicting the global accident scenario in natural circulation reactors ...

International Nuclear Information System (INIS)

The computer programme COMMIX-2 describes steady state and transient multidimensional single- and two-phase fluid flows with heat transfer in nuclear reactor components and multicomponent systems. Originally from the Argonne National Laboratory, the code has been further developed at the Kernforschungszentrum Karlsruhe. The original Point-SOR iterative method for the solution of a Poisson-like equation describing the pressure distribution in the fluid as well as the transport of enthalpy and turbulent quantities has been complemented with iterative and direct line- and block-methods. None of the newly implemented methods is original in itself but their implementation into the computer code, which can describe the most general shapes of definition domains, gave a code speed-up by a factor of 2-5, depending on the problem treated. The code capabilities are assessd ...

International Nuclear Information System (INIS)

The measurement of K(alpha)/K(beta) X-ray intensity ratio indicates the excitation process and the chemical state of the atom examined. Theoretical considerations on the origin of this property are presented. The difference of these ratios following photoionization and electron capture are calculated by a computer code generating transition matrix elements. The theoretical assumptions involved in the codes are tested by comparing the experimental and calculated data concerning 3d transition element atoms. (D.Gy.).

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We prove that the 1984 protocol of Bennett and Brassard (BB84) for quantum key distribution is secure. We first give a key distribution protocol based on entanglement purification, which can be proven secure using methods from Lo and Chau's proof of security for a similar protocol. We then show that the security of this protocol implies the security of BB84. The entanglement purification based protocol uses Calderbank-Shor-Steane codes, and properties of these codes are used to remove the use of quantum computation from the Lo-Chau protocol. (c) 2000 The American Physical Society.

International Nuclear Information System (INIS)

When analyzing the loss-of-coolant accidents at VVER reactor NPP the problem of the effect of noncondensable gases on heat transfer in a horizontal steam generator (HSG) is gaining in importance. Based on the RELAP5/MOD3.2 computer code one analyzed the experiments to condense steam-and-gas mixture in a HSG. The calculations are shown to predict satisfactorily duration of steam generator poisoning from noncondensable gas

Energy Technology Data Exchange (ETDEWEB)

In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode ...

Science.gov (United States)

A computational fluid dynamics (CFD) is coupled with a computational structural dynamics (CSD) to simulate the unsteady rotor flow with aeroelasticity effects. An unstructured upwind Navier-Stokes solver was developed for this simulation, with 2nd order time-accurate dual-time stepping method for temporal discretization and low Mach number preconditioning method. For turbulent flows, both the Spalart-Allmaras and Menter's SST model are available. Mesh deformation is achieved through a fast dynamic grid method called Delaunay graph map method for unsteady flow simulation. The rotor blades are modeled as Hodges & Dowell's nonlinear beams coupled flap-lag-torsion. The rotorcraft computational structural dynamics code employs the 15-dof beam finite element formulation for modeling. The structure code was validated by comparing the natural frequencies of a rotor model with UMARC. The ...

International Nuclear Information System (INIS)

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used ...

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Essential to today's modern refineries and the gasoline production process are fluidized catalytic cracking units. By using a computational fluid dynamics (CFD) code developed at Argonne National Laboratory to simulate the riser, parametric and sensitivity studies were performed to determine the effect of catalyst inlet conditions on the riser hydrodynamics and on the product yields. Simulations were created on the basis of a general riser configuration and operating conditions. The results of this work are indications of riser operating conditions that will maximize specific product yields. The CFD code is a three-dimensional, multiphase, turbulent, reacting flow code with phenomenological models for particle-solid interactions, droplet evaporation, and chemical kinetics. The code has been validated against pressure, particle loading, and product yield measurements. After ...

CERN Document Server

A new N-body and hydrodynamical code, called RAMSES, is presented. It has been designed to study structure formation in the universe with high spatial resolution. The code is based on Adaptive Mesh Refinement (AMR) technique, with a tree based data structure allowing recursive grid refinements on a cell-by-cell basis. The N-body solver is very similar to the one developed for the ART code (Kravtsov et al. 97), with minor differences in the exact implementation. The hydrodynamical solver is based on a second-order Godunov method, a modern shock-capturing scheme known to compute accurately the thermal history of the fluid component. The accuracy of the code is carefully estimated using various test cases, from pure gas dynamical tests to cosmological ones. The specific refinement strategy used in cosmological simulations is described, and potential spurious effects associated to shock ...

Energy Technology Data Exchange (ETDEWEB)

This research aims to develop reliable, advanced system thermal-hydraulic computer code and to quantify the uncertainties of code to introduce the best estimate methodology of ECCS for LBLOCA. Although the one of best estimate code, RELAP5/MOD3.1 was introduced from USNRC, several deficiencies in its reflood model and some improvements have been made. The improvements consist of modification of reflood wall heat transfer package and adjusting the drop size in dispersed flow regime. The tome smoothing of wall vaporization and level tracking model are also added to eliminate the pressure spike and level oscillation. For the verification of improved model and quantification of associated uncertainty, the FLECHT-SEASET data were used and upper limit of uncertainty at 95% confidence level is evaluated. (Author) 30 refs., 49 figs., 2 tabs.

International Nuclear Information System (INIS)

This paper describes the study of particles' dispersion in an isotropic turbulent flow. The particle's motion and the turbulent flow characteristics are calculated independently. While the particles' displacement is computed by the author's code, the flow is simulated with a commercial code : PowerFLOW. The particles and the flow are coupled through the relative velocity component of the aerodynamic force. When the simulated flow is turbulent, a turbulence regeneration model is used in order to get the flow instantaneous velocity. Validation of the method is done by comparing the particles' dispersion obtained with experimental results from literature and with the results calculated by FLUENT. Good accordance is found between numerical studies and experimental results. However, comparison between results of PowerFLOW coupled to the author's code and results from FLUENT shows differences when the ...

International Nuclear Information System (INIS)

To develop the advanced subchannel analysis code, the dominant factors that influence the boiling transitional process must be taken into account in the mechanistic constitutive equations based on the flow geometries and the fluid properties. The dominant factors that influence the boiling transitional processes are (1) the gas-liquid re-distribution by cross flow, (2) the liquid film dryout, (3) the two-phase flow regime transition, (4) the droplet deposition, and (5) the spacer-droplet interaction. At first, we indicated the strategy for the development of the constitutive equations for the five dominant factors based on the experimental database by the latest measurement technique and the latest computational fluid dynamics method. Then, the problems of the present constitutive equations and the improvement plan of the constitutive equations were indicated. Finally, the layered structure for the two-phase/three-field subchannel ...

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This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart ...

Energy Technology Data Exchange (ETDEWEB)

A validated computational fluid dynamics (CFD) computer code, ICRKFLO, was used to investigate the scale-up effects on the coke yields of thermal cracking riser factors. Comparisons were made for calculated coke yields of pilot- and commercial-scales riser units. Computational results show that the riser aspect ratio, reaction temperature, particle residence time, and particle/oil ratio have major impacts on the coke yield. A computational experiment was conducted to determine optimal operating conditions for a conceptual design of a commercial-scale riser unit. This experiment showed that the performance loss in scale-up from pilot to commercial scale may be almost completely recovered through optimizing the operating conditions after scale-up using the CFD simulations as a guide.

International Nuclear Information System (INIS)

Ppercase(femaxi-iv) is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of ppercase(femaxi-iv) is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of ppercase(femaxi-iv) was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in ppercase(femaxi-iv) and some results of applications on the experimental data. ((orig.)).

International Nuclear Information System (INIS)

The SCDAP computer code, which is being developed under the sponsorship of the United States Nuclear Regulatory Commission, models the progression of light water reactor core damage including core heatup, core disruption and debris formation, debris heatup, and debris melting. SCDAP is being used to help identify and understand the phenomena that control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and interpretation of severe fuel damage experiments and data. Comparisons between SCDAP calculations and the experimental data showed good agreement. Calculated and measured bundle temperatures for SFD-ST were within 200 K for the entire bundle and within 20 K for maximum cladding temperatures. For ESSI-2, calculated and measured maximum cladding temperatures were within 50 K, and the extensive liquefaction and relocation that was calculated was in agreement with ...

International Nuclear Information System (INIS)

Liquid sodium is used as coolant in Fast Breeder Reactors (FBR). There is a likelyhood of sodium spillage in ambient air in the Steam Generator Building (SGB) of the FBR plant. Due to high chemical reactivity with oxygen, especially at temperatures greater than 573 K, it catches fire very easily. In order to carryout safety related experimental studies for different modes of sodium fires and to develop suitable mathematical models for the assessment of their consequences, an experimental facility (SFEF, Sodium Fire Experimental Facility) is being setup a IGCAR, Kalpakkam. The SFEF is having a 540 m"3 volume experimental hall. Stainless steel linear will be provided on the inside surfaces of experimental hall walls, ceiling and floor. Analysis has been carried out for enclosed sodium pool fire scenarios in SFEF by using sodium pool fire code SOFIRE II, which estimates the thermal transients like pressure rise, gas temperature rise, cell wall ...

International Nuclear Information System (INIS)

This paper identifies the event sequences that require thermal-hydraulic analyses for the success criteria of probabilistic safety analysis (PSA). The selection of the sequences is performed based on the review of the NEI Peer Review Process Guidance and ASME PRA Standard. Success criteria are the important element of PSA quality. Success criteria decide the success or failure of the key function in the PSA event tree. It is defined as a minimum set of components/trains of system required to mitigate an accident. Thermal-hydraulic codes are generally used to derive time-related criteria in the PSA, such as operator action time used in human reliability analysis (HRA), event timing, and time to recover the component or the power. This paper suggests the use of the MARS code for the T-H analysis to obtain the success criteria and sequence timing, and operator action time. In the Kori Units 3 and 4 PSA report, the T-H analyses ...

International Nuclear Information System (INIS)

With regard to the operation of nuclear power plants, Electricite de France formally recognized in 1973 that it was necessary for safety reasons and economically acceptable to adopt organizational principles of quality assurance that would be applicable both to its own activities and to those of its suppliers. Generally speaking, the form and spirit of the quality assurance programme chosen is based largely on the Code of Practice No. 50-C-QA. In particular, the programme focuses on the flexible character of quality assurance requirements and stresses that in the final analysis product quality depends above all on those to whom the project has been assigned, because it is they who are responsible for meeting the quality objectives set. Ten years of experience with the suppliers of Electricite de France has shown that these suppliers, after some initial difficulty, have been able to adapt the application of quality assurance so as to achieve ...

International Nuclear Information System (INIS)

Core-collapse supernovae are among Nature's most energetic events. They mark the end of massive star evolution and pollute the interstellar medium with the life-enabling ashes of thermonuclear burning. Despite their importance for the evolution of galaxies and life in the universe, the details of the core-collapse supernova explosion mechanism remain in the dark and pose a daunting computational challenge. We outline the multi-dimensional, multi-scale, and multi-physics nature of the core-collapse supernova problem and discuss computational strategies and requirements for its solution. Specifically, we highlight the axisymmetric (2D) radiation-MHD code VULCAN/2D and present results obtained from the first full-2D angle-dependent neutrino radiation-hydrodynamics simulations of the post-core-bounce supernova evolution. We then go on to discuss the new code Zelmani which is based on the open-source HPC ...

Energy Technology Data Exchange (ETDEWEB)

A review of analytical design methods used for predicting reactor core flow and temperature distributions is presented with emphasis on LMFBR's. The paper also briefly describes and contrasts the methods used for LWR's. These methods are global analysis, subchannel analysis, distributed parameter, and hybrid analysis. The evolution of the local and subchannel analysis methods is presented. Data used for code validation are also presented. Current research and development needs are identified and discussed. Areas identified for future research and development include methods and expermental data for analysis of distorted bundles and natural convection. Methods that have been developed for predicting the safety performance of LMFBR's and LWR's are not within the scope of this paper.

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The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only a limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri ...

International Nuclear Information System (INIS)

The present Manual provides guidance and illustrative examples for applying a method by which graded quality assurance requirements may be determined and adapted to the items and services of a nuclear power plant in conformance with the requirements of the IAEA Nuclear Safety Standards (NUSS) Code and Safety Guides on quality assurance. The Manual replaces the previous publication IAEA-TECDOC-303 on the same subject. Various methods of grading quality assurance are available in a number of Member States. During the development of the present Manual it was not considered practical to attempt to resolve the differences between those methods and it was preferred to identify and benefit from the good practices available in all the methods. The method presented in this Manual deals with the aspects of management, documentation, control, verification and administration which affect quality. 1 fig., 4 tabs.

Energy Technology Data Exchange (ETDEWEB)

RADSOLVER is a computer program which calculates the radiation energy transport in cavity type receivers having an arbitrary number of apertures through which collimated beams of solar radiation enter. In contrast to the common assumption of gray (or semi-gray) surfaces used in the modeling of radiation transport, RADSOLVER accounts for the wavelength-dependence of emission, absorption and reflection with a band model of the radiative properties. It is intended that this report serve both as an instruction manual for the use of the RADSOLVER code and a vehicle for presenting the underlying theory. Illustrative examples along with input and output are presented.

Science.gov (United States)

This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate.

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As nuclear research has become more prevalent, environmental contamination from the disposal of radioactive waste has become a prominent issue. At Los Alamos National Laboratory (LANL) in northern New Mexico, radioactive contamination from disposal operations has raised some very specific concerns. Material Disposal Area G (Area G) is the primary low-level radioactive waste disposal site at LANL and occupies an area adjacent to land belonging to the Native American community of the Pueblo of San Ildefonso. Analyses of soil and vegetation collected from the perimeter of Area G have shown concentrations of radionuclides greater than background concentrations established for northern New Mexico. As a result, Pueblo residents had become concerned that contaminants from Area G could enter tribal lands through various ecological pathways. The residents specifically questioned the safety of consuming meat from elk and deer that forage near Area G and then migrate onto ...

International Nuclear Information System (INIS)

Monte Carlo criticality calculations have been performed for over 50 years for reactor physics and criticality safety applications. With today's faster computers, these calculations are being carried out to greater precision (smaller uncertainties) in keff, and detailed distributions of power and reaction rates are being computed routinely. This paper provides a review of the fundamental theory of Monte Carlo criticality calculations, with guidance on practical methods for: (1) assuring convergence of both keff and the source distribution, (2) minimizing the bias in keff and reaction rate distributions, and (3) dealing with the under-prediction bias in uncertainties for keff and reaction rate distributions. (authors)

Energy Technology Data Exchange (ETDEWEB)

With the aid of a DOE grant (No. DE-FC26-01NT41050), Stolar Research Corporation (Stolar) developed the Horizon Sensor (HS) to distinguish between the different layers of a coal seam. Mounted on mining machine cutter drums, HS units can detect or sense the horizon between the coal seam and the roof and floor rock, providing the opportunity to accurately mine the section of the seam most desired. HS also enables accurate cutting of minimum height if that is the operator's objective. Often when cutting is done out-of-seam, the head-positioning function facilitates a fixed mining height to minimize dilution. With this technology, miners can still be at a remote location, yet cut only the clean coal, resulting in a much more efficient overall process. The objectives of this project were to demonstrate the feasibility of horizon sensing on mining machines and demonstrate that Horizon Sensing can allow coal to be cut cleaner and more efficiently. Stolar's primary goal was ...

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The WBURN (2-D, 2-group, coarse mesh) code is developed to analyze the equilibrium core characteristics of CANDU-PHWR. The equilibrium characteristics of Wolsung reactor computed by using WBURN are compared with the values given in the Wolsung FSR. The changes of equilibrium core characteristics caused by the variation of design parameters for operating conditions are also investigated. The numerical results indicate that the average discharge irradiation in the Wolsung reactor can be increased up to about 5%.

International Nuclear Information System (INIS)

To further optimize thermo-hydraulic capability of HCSB-TBM, two different thermo-hydraulic schemes would be proposed and investigated by relevant calculating and simulating of thermo-dynamics in normal and peak heat flux condition. With computational fluid dynamics (CFD) code FLUENT, the actual temperature distribution of key components of HCSB-TBM is simulated. (authors)

Science.gov (United States)

Results are given of numerical simulation of three-dimensional pattern of flow of a two-phase steam-water mixture in the house of a PGV-1000 horizontal steam generator obtained using the BAGIRA best-estimate thermohydrodynamic computer codes. The space distributions of velocities and local void fractions in the steam generator housing for different modes of operation of power-generating unit are calculated and compared with available experimental data.

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Methods of solving Navier-Stokes equations for turbulent channel flow are given. These numerical solutions utilize either Neumann or Dirichlet boundary conditions. Computer codes were written and tested and are discussed. Digital image processing of flow visualization video sequences, taken simultaneously with vorticity probe measurements in a turbulent boundary layer, were carried out. 4 figs. (GHH)

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We present in this paper the computer code BACCHUS, to analyze the thermal-hydraulics in a rod bundle in single or two-phase flow regime. The model is 2-D and uses the porous body approach. The two-phase model is an extension of the classical homogeneous model, and includes a differential non-equilibrium equation. Results are shown for the extension of the boiling region in a 19-pin bundle.

Science.gov (United States)

This paper presents a case study in the design and implementation of a numerical weather prediction model on a supercomputer (CRAY-1). Following a historical introduction to the evolution of the model, the governing equations of the model are presented and the numerical solution of these forecast equations is described. A brief tutorial on the architecture of the CRAY-1 is presented with a discussion of how it affects the choice of algorithms and code design of the model. A summary of the advantages gained by use of the vector aspects of the CRAY-1 is included.

International Nuclear Information System (INIS)

A safety analysis code to design a pressurized water reactor and to obtain the licenses including entire proprietary rights is under development in domestic R and D project. The tasks of KAERI is to develop the constitutive relations including models for defining flow regimes and flow regime related models for inter-phase friction, wall frictions, wall heat transfer, and interphase heat and mass transfer in the two-phase three-field equations. In this paper, the process will be presented for choosing the best flow regime maps which occur in gas-liquid two-phase flow in horizontal and vertical tubes.

Energy Technology Data Exchange (ETDEWEB)

17 papers are presented. Topics covered include: the POG system - a new concept in the use of ANFO; demolition of a motorway bridge; presplit and smooth blasting; VIBReX - a predictive code for assessing the effect of blast design on ground vibration; ground vibrations from blasting; digital seismographs; human response to blasting and the effects on planning conditions; landform construction by restoration blasting; use of small diameter explosives; efficient priming; safety management in the explosives industry; and the law on packaging of explosives. Two papers have been abstracted separately.

International Nuclear Information System (INIS)

A lot of effort has been made to spread the application of RTOS in embedded system development in China. RTOS, #mu#C/OS and #mu#C/OS-II, with its opened source code, is well known in embedded systems. Learning RTOS, Using RTOS, becomes more and more popular. The Book of 'Micro C/OS-II The Real-Time Kennel Second Edition' was published in 2002. Here some new progress of the RTOS, especially the safety certifications, the reliability of software is introduced. (authors)

Energy Technology Data Exchange (ETDEWEB)

This thesis discusses an advanced dynamical two-dimensional cylinder symmetric model for the high temperature part of the carbothermic silicon metal process, and its computer encoding. The situation close to that which is believed to exist around one of three electrodes in full-scale industrial furnaces is modelled. This area comprises a gas filled cavity surrounding the lower tip of the electrode, the metal pool underneath and the lower parts of the materials above. The most important phenomena included are: Heterogeneous chemical reactions taking place in the high-temperature zone (above 1860 {sup o}C), Evaporation and condensation of silicon, Transport of materials by dripping, Turbulent or laminar fluid flow, DC electric arcs, Heat transport by convection, conduction and radiation. The results from the calculations, such as production rates, gas- and temperature distributions, furnace- and particle geometries, fluid flow fields etc, are presented graphically. ...

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The Stockpile Stewardship Program (SSP) is a single, highly integrated technical program for maintaining the safety and reliability of the U.S. nuclear stockpile. The SSP uses past nuclear test data along with future non-nuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production capability to support stockpile requirements. This integrated national program will require the continued use of current facilities and programs along with new experimental facilities and computational enhancements to support these programs. The Advanced Simulation and Computing program (ASC) is a cornerstone of the SSP, providing simulation capabilities and computational resources to support the annual ...

Energy Technology Data Exchange (ETDEWEB)

The 9/11 event in the U.S.A has increased international terror possibilities against nuclear facilities including nuclear power plants(NPPs). It is necessary to assess the performance of existing physical protection system(PPS) at nuclear facilities based on such malevolent acts. A PPS is a complex configuration of detection, delay, and response elements. Several techniques are available to analyze a PPS and evaluate its effectiveness. Sandia National Laboratory(SNL) in the USA was developed a System Analysis of Vulnerability to Intrusion (SAVI) computer code for this purpose. It is powerful software for evaluating the effectiveness of PPS against outsider threats. This study presents the performance assessment of the PPS at Wolsong NPP using SAVI code. The first is that the SAVI constructs the site-specific Adversary Sequence Diagram (ASD) of the PPS. This provides a methods of graphically representing the PPS composed of ...

Energy Technology Data Exchange (ETDEWEB)

A tiered communications architecture for managing network traffic in a distributed system. Communication between client or control computers and a plurality of hardware devices is administered by agent and monitor devices whose activities are coordinated to reduce the number of open channels or sockets. The communications architecture also improves the transparency and scalability of the distributed system by reducing network mapping dependence. The architecture is desirably implemented in a proton beam therapy system to provide flexible security policies which improve patent safety and facilitate system maintenance and development.

International Nuclear Information System (INIS)

Objective: To establish a computer program for rapid calculation of internal committed effective close through inhalation and ingestion pathways under radiological emergency. Methods: Visual Basic 6.0 is used to compile the generic procedures of internal committed effective dose by inhalation and ingestion in IAEA-TECDOC-1162, Generic Procedures for Assessment and Response during a Radiological Emergency. Results: The assessment methodology of internal committed effective dose by inhalation and ingestion under radiological emergency in the report IAEA-TECDOC-1162 can be coded into a computer program. Conclusions: This research provides a rapid method of assessment for internal committed effective dose by inhalation and ingestion under radiological emergency, and it may provide needed dosimetry data for treatment under emergency response. (authors)

Energy Technology Data Exchange (ETDEWEB)

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel ...

International Nuclear Information System (INIS)

Among others, one of the main activities in the Nuclear Engineering and Fluid Mechanics Department of the Engineering School in Bilbao, is the study of liquid metals behavior. And for this purpose the CFD code FLUENT is being used. Currently, the code is being applied to the use of Lead-Bismuth eutectic (LBE) as the coolant of an accelerator driven system (ADS) and also as the target for a neutron source. In this paper, ANSALDO's Energy Amplifier Demonstration Facility is simulated, paying attention only on the coolant. As it will be later explained, natural convection is a very important issue, because the philosophy for safety systems in nuclear devices tends to consider passive technologies. The purpose is to avoid electrical machines like pumps, so the core should remain coolable, even if there is a blackout. To get this natural circulation, heat transfer plays a main role, and as turbulence enhances the heat transfer, ...

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As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known ...

Energy Technology Data Exchange (ETDEWEB)

As the result of 15 years of research (50 staff years of effort) Argonne National Laboratory (ANL), through its involvement in fluidized-bed combustion, magnetohydrodynamics, and a variety of environmental programs, has produced extensive computational fluid dynamics (CFD) software and models to predict the multiphase hydrodynamic and reactive behavior of fluid-solids motions and interactions in complex fluidized-bed reactors (FBRS) and slurry systems. This has resulted in the FLUFIX, IRF, and SLUFIX computer programs. These programs are based on fluid-solids hydrodynamic models and can predict information important to the designer of atmospheric or pressurized bubbling and circulating FBR, fluid catalytic cracking (FCC) and slurry units to guarantee optimum efficiency with minimum release of pollutants into the environment. This latter issue will become of paramount importance with the enactment of the Clean Air Act Amendment (CAAA) of 1995. ...

Energy Technology Data Exchange (ETDEWEB)

Installation of the major hardware subsystems for MFTF is nearing completion. These subsystems include the Fusion Chamber System, the eighty KV Neutral Beam System, the Superconducting Magnet System, and the Personnel Safety System. The Local Controls group has undertaken a uniform aproach to implementing the control systems for all of these hardware subsystems. This approach has two major aspects: (1) to provide a stand-alone computer control system with a remote, portable terminal so that computer control can be provided at the site of the hardware for initial testing, (2) to provide hardware simulators so that the complicated MFTF computer control system can be tested independent of the hardware. The software and hardware tools which were developed to carry out this plan will be described. Our experiences with bringing up subsystems containing up to 900 separate channels of control and status will ...

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A stochastic wind simulation for VAWTs (VSTOC) has been developed which yields turbulent wind-velocity fluctuations for rotationally sampled points. This allows three-component wind-velocity fluctuations to be simulated at specified nodal points on the wind-turbine rotor. A first-order convection scheme is used which accounts for the decrease in streamwise velocity as the flow passes through the wind-turbine rotor. The VSTOC simulation is independent of the particular analytical technique used to predict the aerodynamic and performance characteristics of the turbine. The VSTOC subroutine may be used simply as a subroutine in a particular VAWT prediction code or it may be used as a subroutine in an independent processor. The independent processor is used to interact with a version of the VAWT prediction code which is segmented into deterministic and stochastic modules. Using VSTOC in this fashion is very efficient with regard to decreasing ...

Energy Technology Data Exchange (ETDEWEB)

This paper discusses the application of a Computational Fluid Dynamics (CFD) code (FLUENT) to predict temperature and velocity profiles, and particle trajectories in an industrial fired by a low NO{sub x} pulverized coal swirl burner. The predictions of the code were compared to experimental measurements of gas temperature, particle size distribution, and particle velocities. The diagnostics employed were a suction pyrometer and a Particle Counter Sizer Velocimeter Probe (PCSV-P). Good agreement was found between the temperature predictions and the measurements. In situ particle size distributions were predicted accurately using the Baum and Street model with a swelling factor of zero. The numerical modeling indicated that the Type II flow issued from the burner could be transformed into Type III flow, by eliminating swirl from the primary air stream, which provided the preferred particle trajectories and increased the ...

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Severe accident analysis has been performed for the Wolsung nuclear power= plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks (LOAH) which is caused by loss of off-site power, diesel generators, and DC power. ISAAC (Integrated Severe Accident Analysis Code) computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanisms of calandria vessel and containment. In addition, some insights for accident management program (AMP) are also given. (Author) 5 refs., 1 tab., 12 figs.

International Nuclear Information System (INIS)

The LWR fuel performance analysis computer code, FRAPCON-1, are evaluated to investigate the performance of CANDU fuel elements loaded in Wolsung-1 reactor. The FRAPCON-1 models of neutron flux depression in fuel and of fuel-to-cladding heat transfer are modified, and the validity of fission gas release model for CANDU fuel is evaluated. And the heavy water properties are provided in calculating the heat transfer coefficient between cladding and coolant. By using the modified code, FRAPCON-1-CSK, the sensitivity studies are carried out for Wolsung-1 fuel element design parameters. The performance analysis is also performed for Wolsung-1 fuel elements. The calculated results are discussed in terms of LWR fuel design criteria because of unavailability of CANDU fuel design criteria. (Author).

International Nuclear Information System (INIS)

A nodal integral method is derived for the monoenergetic, steady-state, fixed source neutron diffusion equation in hexagonal geometry based on a coordinate transformation that maps a parallelogram into a rectangle. The new hexagonal nodal diffusion method is implemented in the computer code HND where the discrete-variable equations are solved via an iterative scheme. Because the new method's equations are derived for a rhombus, they can be solved on a sequence of embedded meshes to study the method's error order. Indeed a preliminary numerical error analysis reveals a second-order error in the mesh size, and comparison with finite difference results obtained with the finite difference based BOLD-VENTURE code indicate the superior accuracy of our new nodal method.

Energy Technology Data Exchange (ETDEWEB)

This manual describes the use of the Xyce Parallel Electronic Simulator. Xyce has been designed as a SPICE-compatible, high-performance analog circuit simulator, and has been written to support the simulation needs of the Sandia National Laboratories electrical designers. This development has focused on improving capability over the current state-of-the-art in the following areas: (1) Capability to solve extremely large circuit problems by supporting large-scale parallel computing platforms (up to thousands of processors). Note that this includes support for most popular parallel and serial computers; (2) Improved performance for all numerical kernels (e.g., time integrator, nonlinear and linear solvers) through state-of-the-art algorithms and novel techniques. (3) Device models which are specifically tailored to meet Sandia's needs, including some radiation-aware devices (for Sandia users only); and (4) Object-oriented ...

British Library Electronic Table of Contents (United Kingdom)

A model for risk and reliability analysis of complex multifunctional production process systems is presented. The model employs Monte-Carlo and Markov Chain algorithms that uses a weighted index to train and simulate the fuzzy hazard data sets which represents failure outcomes of risk component transient and non-transient systems. Early simulation results shows that hazard rates and the risk of containment loss from typical floating production and storage offloading (FPSO)-Riser system for the risk components in parallel or series increases exponentially with time and decreases as safety ratings fraction increases. The reliability value decreases with time and safety fraction (SFAC) for all fuzzy hazard classifications. The results of the computed mean time before repair (MTBR) show that t...

British Library Electronic Table of Contents (United Kingdom)

Summary Background There is a need for new artemisinin-based combination therapies that are convenient, effective, and safe. We compared the efficacy and safety of pyronaridine-artesunate with that of artemether-lumefantrine for treatment of uncomplicated P falciparum malaria. Methods This phase 3, parallel-group, double-blind, randomised, non-inferiority trial was undertaken in seven sites in Africa and three sites in southeast Asia. In a double-dummy design, patients aged 3-60 years with uncomplicated P falciparum malaria were randomly assigned in a 2:1 ratio to receive pyronaridine-artesunate once a day or artemether-lumefantrine twice a day, orally for 3 days, plus respective placebo. Randomisation was done by computer-generated randomisation sequence in blocks of nine by study centre....

Energy Technology Data Exchange (ETDEWEB)

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. During the course of follower core assessments, TUV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core ...

Energy Technology Data Exchange (ETDEWEB)

The three-dimensional (3-D), multiphase, computational fluid dynamic (CFD) code FLUENT is used to simulated two-phase flow behaviour in a CANDU header manifold under low (natural circulation) flow conditions. This behaviour was previously inferred from experimental data. The CFD simulations reported here are being used to support these inferences and to obtain a better understanding of phase distribution in the header manifold. The simulations seem to show that the vapor-water mixture models in the FLUENT code do not capture properly phase separation in the header and proper phase branching at the header-feeder connections that have been observed in experiments at low flows. The simulations using discrete-phase model in FLUENT, which tracks the pathlines of the individual vapor bubbles in the water continuum phase, show interesting, complicated and, in some cases, unexpected bubble trajectories from the point of injection ...

Energy Technology Data Exchange (ETDEWEB)

The NUclear Power Engineering Corporation (NUPEC) has developed IMPACT-SAMPSON code to analyze integral behavior of light water nuclear power plants under severe accident conditions. IMPACT-SAMPSON's distinguishing features include interconnected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to severe accident events, and high-speed simulation on parallel processing computers. The integral plant behaviors of typical PWR and BWR under severe accident conditions have been analyzed with the IMPACT-SAMPSON code. The PWR plant analyzed was the three-loop, steel-dry containment type with 2,440 MWt. The AE accident scenario was supposed, that is, LOCA by 6-inch hot leg failure followed by accumulated water injection, but no ECCS and containment spray activation. The BWR plant analyzed was the 3,293 MWt BWR-5, Mark-II containment type. The TQUV accident ...

International Nuclear Information System (INIS)

In this work trailing edge noise levels of a research aerofoil have been computed and compared to aeroacoustic measurements using two different approaches. On the other hand, aerodynamic and aeroacoustic calculations were performed with the full Navier-Stokes CFD code Fluent [Fluent Inc 2005 Fluent 6.2 Users Guide, Lebanon, NH, USA] on the basis of a steady RANS simulation. Aerodynamic characteristics were computed by the aid of various turbulence models. By the combined usage of implemented broadband noise source models, it was tried to isolate and determine the trailing edge noise level. Throughout this work two methods of different computational cost have been tested and quantitative and qualitative results obtained. On the one hand, the semi-empirical noise prediction tool NAFNoise [Moriarty P 2005 NAFNoise User's Guide. Golden, Colorado, July. http://wind.nrel.gov/designcodes/ simulators/NAFNoise] ...

International Nuclear Information System (INIS)

A series of 27 reflood experiments has been carried out in a full-length electrically heated rod bundle facility. The primary objective of these tests was to study the effects of a simultaneous upper plenum and downcomer coolant injection and to provide data for the verification of computer codes. The experimental results indicate that an upper plenum injection alone cools the test rods slowly, a simultaneous coolant injection to the downcomer improves cooling significantly, and a downcomer injection alone cools the test rod bundle best if the total value of the coolant flow rate is the same in these three different cases. If the coolant injected to the upper plenum increases the total flow rate, the quench time of the test rods decreases at all elevations. Quenching time and clad temperature histories calculated with the computer codes NORCOOL-I and FLOOD4 are in a reasonable quantitative agreement ...

International Nuclear Information System (INIS)

The B test series from the course of ISP Nr. 43 was analysed. The boundary conditions measured include the initial temperature of the primary system, the front/slug injection flowrate and temperature, and the pressure drop across the core. Temperature data were collected at 185 thermocouple positions in the downcomer and 38 positions in the lower plenum. The frequency of data acquisition for code predictions was set to 2 Hz, which corresponds to the acquisition frequency of the two experimental setups. Calculations were performed using the FLUENT Computational Fluid Dynamics (CFD) code. This program is based on the finite volume method, The computational mesh was generated by the pre-processor - GAMBIT program. FLUENT uses a control-volume-based technique to convert the governing equations to algebraic equations, which can be solved numerically. This control volume technique consists of integrating the ...

Energy Technology Data Exchange (ETDEWEB)

The present paper is concerned with development and application of a so-called Effective Convection Model (ECM), which aims to provide a detailed, mechanistic description of heat transfer processes in a BWR lower plenum. The ECM is a Computational Fluid Dynamics (CFD)-like tool which employs a simpler and more effective approach to compute heat transfer by solving only energy conservation equation instead of solving the full set of Navier-Stokes and energy equations by a CFD code. We implement the ECM in a CFD code (Fluent), with detailed description of the ECM development, implementation and validation. A dual approach is used to validate the ECM, namely validation against experimental data and against heat transfer results obtained by CFD predictions in the same geometries and conditions. Insights gained from CFD simulations are also used to improve ECM. The ECM capability as an effective tool to ...

British Library Electronic Table of Contents (United Kingdom)

Experimental and computational investigations were performed in order to better understand the mechanical response of rigid targets with various geometries to the detonation of shallow buried explosives. The motion of the targets was measured by use of high-speed digital video photography. This work involved flat targets, targets that were downwardly convex, and targets that were downwardly concave with explosive charges located at various positions beneath the targets. It was observed that, in general, angled hulls - whether downwardly concave or convex - tended to reduce the amount of momentum imparted to the center of mass of the targets. Computations were performed by use of an arbitrary Langrangian-Eulerian treatment in a nonlinear finite element code. A model based on quasi-static te...

Energy Technology Data Exchange (ETDEWEB)

The present paper describes the application of the commercial CFD-code FLUENT 6.1 to the numerical simulation of the flow field in a high specific speed radial fan. Two-dimensional as well as three-dimensional computations are performed using the steady 'Frozen-Rotor method' as well as the unsteady 'Sliding-Mesh method'. The predicted performance curves of the radial fan are compared with the results obtained from measurements on a test stand. Due to the relatively low computation times, the 'Frozen-Rotor method' can be used routinely for design purposes. The results of this approach can be improved if the leakage flow through the gap between the inlet nozzle and the impeller is taken into account. (orig.)

Energy Technology Data Exchange (ETDEWEB)

Ability to predict mechanical response of rock in three dimensions over the spatial and time scales of geologic interest would give the oil and gas industry the ability to reduce risk on prospects, improve pre-project initial reserve estimates, and lower operating costs. A program has recently been initiated, under the auspices of the Advanced Computational Technology Initiative (ACTI), to achieve such a computational technology breakthrough by adapting the unique advanced quasistatic finite element technology developed by Sandia to the mechanics applications important to exploration and production activities within the oil and gas industry. As a precursor to that program, in an effort to evaluate the feasibility of the approach, several complex geologic structures of interest were analyzed with the existing two-dimensional quasistatic finite element code, SANTOS, developed at Sandia. Examples are presented and discussed.

Science.gov (United States)

Three-dimensional computer simulation and analyses of the horizontal steam generator thermal-hydraulics of the WWER 1000 nuclear power plant have been performed for 50% and 75% partial loads, 100% nominal load and 110% over-load. Presented results show water and steam mass flow rate vectors, steam void fraction spatial distribution, recirculation zones, swell level position, water mass inventory on the shell side, and other important thermal-hydraulic parameters. The simulations have been performed with the computer code 3D ANA, based on the 'two-fluid' model approach. Steam-water interface transport processes, as well as tube bundle flow resistance, energy transfer, and steam generation within tube bundles are modelled with {sup c}losure laws{sup .} Applied approach implies non-equilibrium thermal and flow conditions. The model is solved by the control volume procedure, which has been extended in order to ...

Energy Technology Data Exchange (ETDEWEB)

In the field of reactor and fuel cycle physics, particle transport plays and important role. Neutronic design, operation and evaluation calculations of nuclear system make use of large and powerful computer codes. However, current limitations in terms of computer resources make it necessary to introduce simplifications and approximations in order to keep calculation time and cost within reasonable limits. Two different types of methods are available in these codes. The first one is the deterministic method, which is applicable in most practical cases but requires approximations. The other method is the Monte Carlo method, which does not make these approximations but which generally requires exceedingly long running times. The main motivation of this work is to investigate the possibility of a combined use of the two methods in such a way as to retain their advantages while avoiding their drawbacks. Our ...

Energy Technology Data Exchange (ETDEWEB)

Lining shafts and placing them in a given strata needs extensive computation. Therefore, it is natural that the use of electronic computing and of numerical methods in this mining area started early. For shafts being sunk in non-stable water-bearing strata the deformations and stresses of the freezing wall are analysed. The complex system of an outer lining in the shaft sinking stage is designed with regard to the elastic support of the strata. In cases of strata-embedded systems the tubbing liner can be shaped as the loads require. Here analysis of safety against buckling is very important. If a sliding shaft is built as a composite construction, there are advantages in determining the distribution of the material by numerical methods. The analysis of stability is possible only by numerical methods and could not be solved otherwise. The Finite Element Method proves itself as a very general tool to solve the problem of ...

International Nuclear Information System (INIS)

The following are reported: colorimetric determination of Ti in Cr, decontamination of metals by anodic dissolution, computer code for x-ray peak identification, gel-permeation chromatography of CoRezyn 1664 polyester resin, infrared monitoring of gases and vapors, particle size analysis of dusts, nondestructive analysis of Oralloy for removable alpha contamination, dissolution (digestion) of high-fired oxides, assay of U in a Pu-U alloy, measurement of oxide thickness on Pu metal by x-ray diffraction, analysis of precipitated Pu peroxide, differentiation of Dowex SBR and 11 resins, and determination of Pu solubility in simulated lung fluid.

CERN Document Server

The whole computer hardware industry embraced multicores. For these machines, the extreme optimisation of sequential algorithms is no longer sufficient to squeeze the real machine power, which can be only exploited via thread-level parallelism. Decision tree algorithms exhibit natural concurrency that makes them suitable to be parallelised. This paper presents an approach for easy-yet-efficient porting of an implementation of the C4.5 algorithm on multicores. The parallel porting requires minimal changes to the original sequential code, and it is able to exploit up to 7X speedup on an Intel dual-quad core machine.

Energy Technology Data Exchange (ETDEWEB)

This paper presents a numerical simulation with the EUROPLEXUS fast dynamics software of a pipe whipping phenomenon occurring in the thermal hydraulic conditions of a loss of coolant accident in a PWR primary circuit. Different physical phenomena take place simultaneously during the rupture and the whipping of the pipe such as plasticity, contact, large displacements, two-phase flow regime and fluid structure interaction. Two kinds of numerical models - a simplified pipeline model and a mixed 1D/3D model - are considered and compared throughout modelling and computation. Numerical results are compared with experimental data belonging to the Aquitaine II test campaign.

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A neutron collimator system is being designed as part of a neutron imaging facility for computed tomography and real-time neutron radiography research at the through beam port of the University of Texas TRIGA reactor. Lack of sufficient information about collimator systems in a through port from the literature necessitated the use of Monte Carlo calculations using the MCNP code 3 to search for optimal design configuration and materials that maximize the thermal neutron intensity at the image plane while minimizing the fast neutrons and gamma radiation.

CERN Document Server

This report describes studies performed in the framework of the Collimation Task Force organized to support the work of the second International Linear Collider Technical Review Committee. The post-linac beam-collimation systems in the TESLA, JLC/NLC and CLIC linear-collider designs are compared using the same computer code under the same assumptions. Their performance is quantified in terms of beam-halo and synchrotron-radiation collimation efficiency. The performance of the current designs varies across projects, and does not always meet the original design goals. But these comparisons suggest that achieving the required performance in a future linear collider is feasible.

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We apply and modify the computer codes CAIN(P.Chen, G.Horton-Smith, T.Ohgaki, A.W.Weidemann, K.Yokoya, contributed to Workshop on Gamma-Gamma Colliders, Berkeley, CA, March 28-31, 1994; SLAC-PUB-6583.) and GUINEAPIG(D.Schulte (DESY), unpublished.) to make detailed studies of the beam-beam interaction in the SLAC design for a future linear collider (NLC), as well as to the proposed SLC2000 project. Using realistic beam distributions, we present simulations related to the tuning and optimization of luminosity in SLC2000 and NLC.

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The stability-enchancing two-step (SETS) method has been used successfully in the Transient Reactor Analysis Code (TRAC) for several years. The method consists of a basic semi-implicit step combined with a stabilizer step that, taken together, eliminate the material Courant stability limit associated with standard semi-implicit numerical methods. This approach toward stability requires significantly fewer computational operations than a fully implicit method, but currently maintains the first-order accuracy in space and time of its semi-implicit predecessors.

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Computational fluid dynamics(CFD) code FLUENT was used to simulate the thermal hydraulic processes occurring in conceptual design of the accelerator-driven subcritical reactor(ADSR) liquid lead target. The purpose of the analysis is to investigate the thermal hydraulic characteristics of liquid lead as ADSR target material with various target geometries and injection locations of proton beam. In the calculation analysis, the local temperature of the liquid lead target rises to the boiling temperature very rapidly. When the proton beam is injected from the bottom of the target system, the duration time to reach the boiling temperature is longer and the temperature distribution is flatter than other cases.

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Monte Carlo methods have been used to compute k{sub eff} and the fundamental mode eigenfunction of critical systems since the 1950s. While such calculations have become routine using standard codes such as MCNP and SCALE/KENO, there still remain 3 concerns that must be addressed to perform calculations correctly: convergence of k{sub eff} and the fission distribution, bias in k{sub eff} and tally results, and bias in statistics on tally results. This paper provides a review of the fundamental problems inherent in Monte Carlo criticality calculations. To provide guidance to practitioners, suggested best practices for avoiding these problems are discussed and illustrated by examples.

International Nuclear Information System (INIS)

This document is furnished as part of the effort to develop NRC Class 1E Digital Computer Systems Guidelines which is Task 8 of USAF Rome Laboratories Contract F30602-89-D-0100. The report addresses four major topics, namely, computer programming languages, software design and development, software testing and fault tolerance and fault avoidance. The topics are intended as stepping stones leading to a Draft Regulatory Guide document. As part of this task a small scale survey of software fault avoidance and fault tolerance practices was conducted among vendors of nuclear safety related systems and among agencies that develop software for other applications demanding very high reliability. The findings of the present report are in part based on the survey and in part on review of software literature relating to nuclear and other critical installations, as well as on the authors' experience in these areas.

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Currently, x-ray mammography is the method of choice in breast cancer screening programmes. As the mammography technology moves from 2D imaging modalities to 3D, conventional computational phantoms do not have sufficient detail to support the studies of these advanced imaging systems. Studies of these 3D imaging systems call for a realistic and sophisticated computational model of the breast. DeBRa (Detailed Breast model for Radiological studies) is the most advanced, detailed, 3D computational model of the breast developed recently for breast imaging studies. A DeBRa phantom can be constructed to model a compressed breast, as in film/screen, digital mammography and digital breast tomosynthesis studies, or a non-compressed breast as in positron emission mammography and breast CT studies. Both the cranial-caudal and mediolateral oblique views can be modelled. The anatomical details inside the phantom include the lactiferous ...

International Nuclear Information System (INIS)

The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYSR and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are ...

Energy Technology Data Exchange (ETDEWEB)

Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high {sup 235}U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used ...

Energy Technology Data Exchange (ETDEWEB)

The purpose of this paper is to compare the techniques and results obtained when a general industrial or commercial building power electrical system is prepared using the US National Electrical Code{copyright} (NEC) and the International Electrotechnical Commission (IEC) Standards. The final design offer the reader the possibility of making an initial comparison and analyzation of the two methods based of their overall effectiveness (sizing, protection, co-ordination of protective devices with components etc.), safety and cost. In order to make such a comparison a specific example was developed and limited in detail to avoid complexity in this study. The power source selected will be a delta-wye utility supplied transformer feeding various load groups such as lighting and motor circuits (possibly through associated transformers). Relevant NE Code references, IEC standards, and appropriate justifications will be made at ...

International Nuclear Information System (INIS)

The SMART (System-integrated Modular Advanced ReacTor) which is a 330 MWt advanced integral PWR was developed by the KAERI (Korea Atomic Energy Institute) for electricity generation and seawater desalination. A thermal hydraulic evaluation and analysis of the SMART is performed by the TASS /SMR-S (Transient And Setpoint Simulation/System integrated Modular Reactor-Safety). The TASS/SMR-S code has various models reflecting the design features of the SMART such as the drift flux model, the core models (core power and core heat transfer model), the component models, and the specific models. One of the core models is the core heat transfer model. The role of this model is to calculate the heat flux and radial temperature profiles at a fuel rod surface using the relevant heat transfer correlations for all of the heat transfer modes. Also it is modeled to meet the requirements of the 10 CFR 50 appendix K EM model for the CHF (Critical Heat Flux) and ...

International Nuclear Information System (INIS)

Purpose/Objective: With the advent of computed tomography and magnetic resonance imaging, the three dimensional representation of the patient anatomy has become an invaluable resource for better diagnosis and delineation of the target volume and sensitive structures in radiation therapy. Although the therapeutic linear accelerator industry has made available highly sophisticated equipment, the aggressiveness in dose prescription and delivery has to be complimented by accurate dose computation methods. We have adopted a convolution/superposition algorithm for the calculation of absolute dose that fully accounts for the external shape and internal structure of the patient for photon treatment radiotherapy. In this paper, we will discuss the principles of the convolution algorithm and we will show how the computed dose compares to clinically relevant treatment techniques. Materials and Methods: A computer ...

International Nuclear Information System (INIS)

Loading pattern optimization (LPO) for a PWR in nuclear power plant contains three parts: fuel assembly location optimization, burnable poison placement optimization, and used fuel assembly orientation optimization. To solve the former two parts, this paper devises an innovative stochastic evolutionary algorithm-Interval Bound Algorithm (IBA), which can optimize fuel assembly location and burnable poison placement together. IBA just uses the fuel assembly's infinite multiplication factor to get rid of unfavorable patterns and to explore new promising solution space. To solve the last part, this paper applies Estimation of Distribution Algorithms (EDAs), which also belong to evolutionary algorithms. These three parts depend on each other, so it is better not to solve them separately. In order to optimize these parts in a coupled way, we use Symbiotic Co-evolutionary Algorithm (SCA) to incorporate IBA and EDAs. This technique could reflect the real optimization process. Based on these ...

Energy Technology Data Exchange (ETDEWEB)

Results obtained on induced current and voltage caused by a high voltage transmission lines running parallel at a trolley bus contact lines, are presented. A computer program in Basic language, was developed to calculate the electric induction, problems of safety, a theoretical development and mathematical model of the electric factors between two lines (inductor and induced) are also discussed. 11 figs., 1 tab., 8 refs.

Energy Technology Data Exchange (ETDEWEB)

In nuclear reactor safety research, the question of the possible consequences of delayed core reflood during severe accidents or anticipated transient without scram transients in boiling water reactors (BWRs) has been raised. One can envisage a very low probability accident scenario leading to core uncovery and core heat-up, followed by control blade melting and subsequential delayed reflooding of the core with unborated water before its degradation. Reflooding of the hot core causes significant increases in the peak heating, melting, and hydrogen production rates, thus increasing the probability of core degradation. However, as has been established, debris beds formed from shattered fuel rods and quenched fuel melt will be undermoderated. The reflood process of a voided, intact core was examined using the TRAC/BFI CODE.

British Library Electronic Table of Contents (United Kingdom)

To form a licensing basis for a new methodology for a fuel channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown fuel channel analysis has been developed, and tested for a high temperature thermal-chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal-chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results ...

International Nuclear Information System (INIS)

The Cycle 8A static tests conducted in the Fast Flux Test Facility (FFTF) during 1986 have resulted in the separation of various feedback reactivity components. These feedback components, described by closed-form equations depending only on the reactor temperature field, can be regarded as database for the validation and/or calibration of feedback mechanistic models. The SASSYS safety analysis code contains the most developed feedback reactivity models and was selected for the comparison study between database and mechanistic calculations for the FFTF. Although detailed feedback models for control rod repositioning and core radial expansion/bowing exist, only the simple models were available in SASSYS at the time of this study. The results are described in this paper.

International Nuclear Information System (INIS)

Within the United Kingdom, the regulatory body having responsibility for the licensing of nuclear installations is the Health and Safety Executive (HSE). The Nuclear Installations Inspectorate (NII) is that part of HSE which administers this function. Discussions on the applicability of quality assurance (QA) to licensed sites began in 1974, and an internal report was published in 1975. In parallel with work going on at the International Atomic Energy Agency (IAEA) to prepare Quality Assurance for Safety in Nuclear Power Plants: A Code of Practice, Safety Series No. 50-C-QA, NII published a second report in 1978 entitled A Guide to the Quality Assurance Programme for Nuclear Power Plants. In 1980, the construction of advanced gas cooled reactors at Heysham 2 and at Torness was licensed, and a condition was attached to the licences requiring the licensees to submit their QA arrangements to the NII for ...

Science.gov (United States)

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Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: During calculations of SAR accidents and transients it is necessary to perform steam generator simulation. Best accuracy is 3D transient calculations presented in report. Main outcomes of work was next: 1. There was shown by analysis the applicability of code TRAC (Los-Alamos laboratory) for thermal - hydraulic calculations of horizontal steam generator PGV-1000M. Special nodalization scheme was developed for it purposes. 2. Validation and selection of thermal-hydraulic correlations for improvement of using the code at calculation PGV-1000M were performed. As result Labuntsov formula is recommended for horizontal SG. 3. Calculations of nominal mode operation of PGV-1000M for cross-verification with code STEG (Electrogorsk Research and Engineering Center EREC) during its verification were performed. Solution by TRAC was obtained for transient problem after stabilization time. 4. Development ...

Energy Technology Data Exchange (ETDEWEB)

Two computational problems were worked on for this study. The first chapter examines the option of coal combustion using oxygen feed with carbon dioxide recycle to control the adiabatic flame temperature. Computer simulations using an existing state-of-the-art 3-dimensional computer code for turbulent reacting flows with reacting particles were employed to study the effects of increased carbon dioxide mole fraction on the char burnout, radiant heat transfer, metal partitioning, and NOx formation. The second chapter compares assumptions for the CO/CO{sub 2} ratio at the surface of mineral inclusions made in previous studies to predictions obtained from a pseudo-steady state kinetic model (SKIPPY) for a single porous particle. The detailed kinetic simulations from SKIPPY for varying particle sizes and bulk gas compositions were used to develop algebraic expressions for the CO/CO{sub 2} ratio that can be ...

International Nuclear Information System (INIS)

Experimental and computational studies were performed to confirm the enhancement of turbulent heat transfer performance in the 6x6 simulated rod bundle subchannel by generating the large scale secondary vortex flow. Experimental studies were carried out at Reynolds Number 10,000 with atmospheric condition. Axial variations of mean velocity and turbulent intensity in the rod bundle subchannel were measured by the 2-color LDV measurement system. Computational works were accomplished using the commercial CFD code, FLUENT. Normal operating condition of Uljin 3, 4 nuclear power plant is used in computation works as an initial condition. The LSSVF mixing vanes generate the most strong secondary flow vortices that maintain about 35 D_H after the spacer grid. The LSSVF mixing vane influences strongly to flow mixing in adjacent subchannels because large scale stream wise vortices in subchannel sustain two times ...

International Nuclear Information System (INIS)

Development in Instrumentation and Control (I and C) technology provides more convenience and better performance, thus, adopted in many fields. To adopt newly developed technology, nuclear industry requires rigorous V and V procedure and tests to assure reliable operation. Adoption of digital system requires verification and testing of the OS for licensing. Commercial real-time operating system (RTOS) is targeted to apply to various, unpredictable needs, which makes it difficult to verify. For this reason, simple, application-oriented realtime OS is developed for the nuclear application. In this work, we show how to verify the developed RTOS at each development lifecycle. Commercial formal tool is used in specification and verification of the system. Based on the developed model, software in C language is automatically generated. Tests are performed for two purposes; one is to identify consistency between the verified model and the generated code, the other is to ...

Energy Technology Data Exchange (ETDEWEB)

In this article two integral computational fluid dynamics methods for steady-state and transient vehicle aerodynamic simulations are described using a Chevrolet Corvette ZR-1 surface panel model. In the last decade, road-vehicle aerodynamics have become an important design consideration. Originally, the design of low-drag shapes was given high priority due to worldwide fuel shortages that occurred in the mid-seventies. More recently, there has been increased interest in the role aerodynamics play in vehicle stability and passenger safety. Consequently, transient aerodynamics and the aerodynamics of vehicle in yaw have become important issues at the design stage. While there has been tremendous progress in Navier-Stokes methodology in the last few years, the physics of bluff-body aerodynamics are still very difficult to model correctly. Moreover, the computational effort to perform Navier-Stokes simulations from the ...

Energy Technology Data Exchange (ETDEWEB)

MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The ...

Science.gov (United States)

Under support from the French Space Agency (CNES), a 3D+t dusty-gas model of Comet 67P/Churyumov-Gerasimenko is being developed, to compute, from the first 2014 Rosetta orbital data, the aerodynamic forces exerted on the Rosetta orbiter and on the descent lander. We report the recently developed dust dynamics part of the code. The multi-species (presently H2O and CO) gas code is optimized in terms of computational speed owing to the use of two complementary methods: (a) 3D+t Direct Simulation Monte Carlo (DSMC) runs in the non-equilibrium regions adjacent to the surface and very distant from it, and (b) solutions of the Navier-Stokes equations in-between. The model is used presently using Lamy et al. (Space Sci. Rev., 2007, 128, 23) coarse information on 67P nucleus shape and rotation, and a range of possible gas production rates Q for the early Rosetta observations at rh 3 AU (Q 1026 - 1027 s-1). In ...

International Nuclear Information System (INIS)

Recent advances in self-amplified spontaneous emission (SASE) experiments stimulate interest in quantitative comparison of measurements with theory. In this paper we show that the widely used simulation code TDA3D, developed by Tran and Wurtele [Comput. Phys. Commun. 54, 263 (1989)] even though a single frequency code, can be used to determine the output power in the SASE process with excellent approximation in the exponential growth regime. The method applies when the gain is not very high, which is a special advantage, because when the gain is not very high, the analytical calculation is particularly difficult since the exponential growing term does not dominate. The analysis utilizes a scaling relation between the output power and the number of simulation particles in the code TDA3D: left-angle P right-angle=N_#lambda#"'/N_#lambda# left-angle P"'right-angle, where left-angle P right-angle is the ...

International Nuclear Information System (INIS)

In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation ...

International Nuclear Information System (INIS)