WorldWideScience

Sample records for accident codes applications

  1. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jang, S. H.; Chun, S. W.; Jang, H. S. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1993-01-15

    As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-15

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

  6. Application of COREMELT-3D code at analysis of severe fast reactor accidents

    International Nuclear Information System (INIS)

    The code COREMELT for calculations of initial and transition stages of severe accident is considered. It is used to conduct connected calculations of nonstationary neutronic and thermohydraulic processes in sodium fast reactor core. The code has some versions depending on dimensions of solving problem and consists of thermohydraulic module COREMELT and neutronic module RADAR. Using the code COREMELT-3D connected calculations of core disassembly accidents of ULOF and UTOP type have been conducted for sodium fast reactors safety analysis. The main problem of code COREMELT-3D use is duration of calculation, speeding of the code is possible when calculating algorithms are parallelized

  7. Accident consequence assessment code development

    International Nuclear Information System (INIS)

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

  8. Application of the integral code MELCOR for German NPPs and use within accident management and PSA projects

    International Nuclear Information System (INIS)

    The paper summarizes the application of MELCOR to German NPPS with PWR and BWR. A development of different code systems like ATHLET/ATHLET-CD, COCOSYS and ASTEC is done as well at GRS but it is not discussed in this paper. GRS has been using MELCOR since 1990 for real plant calculations. The results of MELCOR analyses are used mainly in PSA level 2 studies and in Accident Management projects for both types of NPPs. MELCOR has been a very useful and robust tool for these analyses. The calculations performed within the PSA level 2 studies for both types of German NPPs have shown that typical severe accident scenarios are characterized by several phases and that the consideration of plant specifics are important not only for realistic source term calculations. An overview of typically severe accident phases together with main accident management measures installed in German NPPs is presented in the paper. Several severe accident sequences have been calculated for both reactor types and some detailed nodalisation studies and code to code comparisons have been prepared in the past, to prove the developed core, reactor circuit and containment/building nodalisation schemes. Together with the compilation of the MELCOR data set, the qualification of the nodalisation schemes has been pursued with comparative calculations with detailed GRS codes for selected phases of severe accidents. The results of these comparative analyses showed in most of the areas a good agreement of essential parameters and of the general description of the plant behaviour during the accident progression. The in general detail of the German plant nodalisation schemes developed for MELCOR contributes significantly to this good agreement between integral and detailed code results. The implementation of MELCOR into the GRS simulator ATLAS was very important for the assessment of the results, not only due to the great detail of the nodalisation schemes used. It is used for training of severe accident

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

    International Nuclear Information System (INIS)

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

  10. Upgrade of a fusion accident analysis code and its application to a comparative study of seven fusion reactor designs

    International Nuclear Information System (INIS)

    Fusion energy has the potential to be a safe and environmentally favorable energy source. The importance of safety necessitates the existence of a computer code having the capability of assessing off-site impacts resulting from postulated fusion reactor accidents. The FUSCRAC3 computer code has been developed for this purpose. FUSCRAC3 calculates doses resulting from inhalation, groundshine, and cloudshine for 259 isotopes as well as doses resulting from ingestion for 145 isotopes. FUSCRAC3's data base includes the most up-to-date dose conversion factors for all four exposure pathways as well as the most current environmental transfer factors for the ingestion pathway. This work presents a detailed description of the modifications made to the existing fusion reactor accident code, FUSCRAC2, in the development of the more advanced FUSCRAC3 computer code. Also included is a report of the validation procedures. Finally, the improved computer code was applied in two ways: to provide a general data base presenting rem per curie data for each isotope and to assess the doses resulting from possible releases from the reactors evaluated in the ESECOM study. Regarding the latter application, it was found that the general trends established in the original study remained unchanged. However, it was determined that the inclusion of the ingestion pathway substantially affects the overall chronic dose. Isotopes of particular interest due to the ingestion contribution include H-3, Ca-45, Fe-55, and Po-210. 12 refs., 2 figs., 12 tabs

  11. Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

  12. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-02-01

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

  13. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

  16. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-02-01

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

  17. The Application of Paret/ANL Code for Accident Analysis on Inadvertent Control Rod Withdrawal for RSG GAS Reactor

    International Nuclear Information System (INIS)

    The analysis is intended to take a look the condition of safety parameters such as fuel and clad temperature, and minimum safety margin against flow instability (S) in the occurrence of inadvertent control rod withdrawal at nominal power, which is performed by PARET/ANL Code. The accident is initiated when all control rods are simultaneously withdrawn with maximum speed of 0.0564 cm/s which consequently gives maximum reactivity insertion rate σρ/σt = 2.82 x 10-4/s, resulting in the Reactor Protection System (RPS) respond to scram the reactor by dropping the control rods into the core. The primary cooling system is assumed to be in normal operation. It is postulated that the first trip signal from over power is not effective to scram the reactor, but only the second signal from Floating Limit Value eventually causes a scram with 0.5 s delays. During the occurrence of inadvertent control rods withdrawal at 30 MW of initial power, the maximum fuel and clad temperature reach 181.29oC and 137.62oC, respectively and the peak power of 37.11 MW. Meanwhile the minimum value of S reaches 2.62. Therefore, during the occurrence of control rods withdrawal at initial power of 30 MW, the integrity of fuel and clad can be maintained secure since they do not exceed the maximum limit of fuel and clad temperature of 207oC and 145oC, respectively and the minimum value of S is still higher than the design limit of 1.48 for anticipated transient

  18. Code strategy for simulating Severe Accident Scenario

    International Nuclear Information System (INIS)

    Severe accident scenarios of Sodium-cooled fast reactors involves various phenomena: core degradation, melt progression towards the core catcher, corium behaviour on the core catcher, energetic corium/sodium interactions, structure mechanical behaviour during expansion phase, containment behaviour, and fission production release and transport. In order to simulate the complete accident scenarios, CEA strategy relies on two sets of calculation codes: a reference set of codes and a set of simplified coupled models dedicated to Probabilistic Risk Assessment analyses. Concerning the reference set, that includes SAS-SFR, SIMMER, CONTAIN, EUROPLEXUS, and TOLBIAC, CEA started, with JAEA and KIT, a validation process based on existing experimental results such as CABRI and SCARABEE programs, and recently against the EAGLE1&2 program results, in the frame of a specific contract with JAEA. Furthermore, CEA is preparing additional experimental programs including in-pile experiments in IGR (NNC reactor), and out-of-pile experiments in the future experimental FOURNAISE facility to be built in CEA Cadarache (France). (author)

  19. Adjoint-based sensitivity analysis for reactor accident codes

    International Nuclear Information System (INIS)

    This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and applications to consider problem/model discontinuities in a general fashion, provide for response (R) formulations required by reactor safety applications, and provide a scheme for accurately handling extremely time-sensitive reactor accident responses. The scheme involves partitioning (dividing) the model into submodels (with spearate defining equations and initial conditions) at the location of discontinuity. Successful partitioning moves the problem dependence on the discontinuity location from the whole model system equations to the initial conditions of the second submodel

  20. Improvement and verification of steam explosion models and codes for application to accident scenarios in light water reactors

    OpenAIRE

    Vujic, Zoran

    2008-01-01

    Steam explosions can occur during an accident with core melting in Light Water Reactors (LWR) as a consequence of the interaction between molten core material with the water inside the Reactor Pressure Vessel (RPV) or, if RPV failure cannot be excluded, due to the release of melt from the RPV into water in the cavity. Generally, steam explosions progresses through two distinct phases, characterized by different time scales for the dominant processes i.e. the premixing and explosion phase. ...

  1. Improvement of Severe Accident Analysis Computer Code and Development of Accident Management Guidance for Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Ko Ryu; Kim, Dong Ha; Kim, See Darl; Song, Yong Mann; Choi, Young; Jin, Young Ho

    2005-03-15

    The objective of the project is to develop a generic severe accident management guidance(SAMG) applicable to Korean PHWR and the objective of this 3 year continued phase is to construct a base of the generic SAMG. Another objective is to improve a domestic computer code, ISAAC (Integrated Severe Accident Analysis code for CANDU), which still has many deficiencies to be improved in order to apply for the SAMG development. The scope and contents performed in this Phase-2 are as follows: The characteristics of major design and operation for the domestic Wolsong NPP are analyzed from the severe accident aspects. On the basis, preliminary strategies for SAM of PHWR are selected. The information needed for SAM and the methods to get that information are analyzed. Both the individual strategies applicable for accident mitigation under PHWR severe accident conditions and the technical background for those strategies are developed. A new version of ISAAC 2.0 has been developed after analyzing and modifying the existing models of ISAAC 1.0. The general SAMG applicable for PHWRs confirms severe accident management techniques for emergencies, provides the base technique to develop the plant specific SAMG by utility company and finally contributes to the public safety enhancement as a NPP safety assuring step. The ISAAC code will be used inevitably for the PSA, living PSA, severe accident analysis, SAM program development and operator training in PHWR.

  2. Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

    International Nuclear Information System (INIS)

    In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)

  3. Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

    Energy Technology Data Exchange (ETDEWEB)

    Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others

    1997-07-01

    In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

  4. Review of Severe Accident Phenomena in LWR and Related Severe Accident Analysis Codes

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

    Full Text Available Firstly, importance of severe accident provision is highlighted in view of Fukushima Daiichi accident. Then, extensive review of the past researches on severe accident phenomena in LWR is presented within this study. Various complexes, physicochemical and radiological phenomena take place during various stages of the severe accidents of Light Water Reactor (LWR plants. The review deals with progression of the severe accidents phenomena by dividing into core degradation phenomena in reactor vessel and post core melt phenomena in the containment. The development of various computer codes to analyze these severe accidents phenomena is also summarized in the review. Lastly, the need of international activity is stressed to assemble various severe accidents related knowledge systematically from research organs and compile them on the open knowledge base via the internet to be available worldwide.

  5. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

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

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Ronaldo Celem

    2001-10-15

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

  7. Severe accident analysis code Sampson for impact project

    Energy Technology Data Exchange (ETDEWEB)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh [Nuclear Power Engineering Corporation, Advanced Simulation Systems Dept., Tokyo (Japan)

    2001-07-01

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

  8. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  9. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  10. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

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

  11. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  12. Application of Multi-physical Coupling Method in Development of Severe Accident Simulation Code SimSA%多物理耦合方法在严重事故仿真软件SimSA开发中的应用

    Institute of Scientific and Technical Information of China (English)

    魏巍; 齐克林; 林旭升; 杨森权; 谭超

    2016-01-01

    In order to expanding the scope of the full scale simulator (FSS ) to severe accident ,a simulation code was developed and named SimSA ,and the main processes of severe accident can be modeled in this code .There are three main modules in the SimSA code ,including thermal-hydraulics module (Therm) ,core behavior module (Core) and containment module (Cont ) . A multi-physical coupling method similar to SCDAP/RELAP5 was used in the integration between Therm and Core .This paper introduced the application process of the multi-physical coupling method in the development of Sim-SA ,and it was used to calculate and analyze the severe accident sequences of loss of coolant (LOCA) with failure of safety injection and station blackout (SBO) with loss of auxiliary feed water (AFW) .The calculation results of this code were compared with the calculation results of MAAP4 code .The results indicate that the application of the multi-physical coupling method in SimSA is successful .%为满足核电厂全范围模拟机对严重事故过程仿真的需求,自主开发了严重事故仿真软件SimSA,能模拟从设计基准事故到严重事故的主要事故过程,并能准确给出相关进程的计算结果.SimSA包含3大主要模块:热工水力模块(Therm)、堆芯行为模块(Core)以及安全壳行为模块(Cont).其中,Therm与Core两个模块的耦合过程中采用了SCDAP/RELAP5相似的基于过程机理的耦合方法.本文结合SimSA软件的具体情况介绍了这种耦合方法的实现过程,并采用耦合后的程序对大破口叠加安注失效及全厂断电叠加辅助给水丧失两个典型初因事故导致的严重事故序列进行了计算,将计算结果与相同初始条件下MAAP4的计算结果进行对比分析.结果表明,SimSA中采用的这种耦合方式是成功的.

  13. EAC european accident code. A modular system of computer programs to simulate LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat transfer together with coolant thermohydraulics in single- and two-phase flow. Temperature variations in fuel, coolant and neighbouring structures induce, in fact, thermal reactivity feedbacks which are added up and put in the neutronics calculation to predict the neutron flux and the subsequent heat generation in the reactor. At this point a whole-core analysis code is necessary to examine for any hypothetical transient whether the various feedbacks result effectively in a negative balance, which is the basis condition to ensure stability and safety. The European Accident Code (EAC), developed at the Joint Research Centre of the CEC at Ispra (Italy), fulfills this objective. It is a modular informatics structure (quasi 2-D multichannel approach) aimed at collecting stand-alone computer codes of neutronics, fuel pin mechanics and hydrodynamics, developed both in national laboratories and in the JRC itself. EAC makes these modules interact with each other and produces results for these hypothetical accidents in terms of core damage and total energy release. 10 refs

  14. DOE modifications to the MAAP [Modular Accident Analysis Program] code

    International Nuclear Information System (INIS)

    This report presents an enhanced model for the MAAP code that addresses fuel-cladding interaction and core mass relocation during core degradation. The main purpose of this work is to assess the potential for in-vessel hydrogen production and to reduce the uncertainty in fission product source term evaluation. The model provides a description of fuel behavior in which the fuel comprises uranium dioxide, zirconium dioxide, and U-Zr-O compounds. The composition of the U-Zr-O compounds and their solidus and liquidus temperatures are calculated throughout the core melt transient. The interaction of control rod materials with fuel and cladding and the relocation of control rod materials are not addressed in this enhanced model. The enhanced core melt progression model has been applied to a hypothetical station blackout accident with a small break via the reactor coolant pump seals. The new model has been benchmarked against both the LOFT experiment LP-FP-2 and the TMI-2 accident prior to the B-loop pump restart. Although some uncertainties and deviations were seen, general agreement was obtained with the experimental data and with the TMI-2 accident. 21 refs., 30 figs

  15. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  16. Dosimetric reconstruction of radiological accident by numerical simulations by means associating an anthropomorphic model and a Monte Carlo computation code

    International Nuclear Information System (INIS)

    After a description of the context of radiological accidents (definition, history, context, exposure types, associated clinic symptoms of irradiation and contamination, medical treatment, return on experience) and a presentation of dose assessment in the case of external exposure (clinic, biological and physical dosimetry), this research thesis describes the principles of numerical reconstruction of a radiological accident, presents some computation codes (Monte Carlo code, MCNPX code) and the SESAME tool, and reports an application to an actual case (an accident which occurred in Equator in April 2009). The next part reports the developments performed to modify the posture of voxelized phantoms and the experimental and numerical validations. The last part reports a feasibility study for the reconstruction of radiological accidents occurring in external radiotherapy. This work is based on a Monte Carlo simulation of a linear accelerator, with the aim of identifying the most relevant parameters to be implemented in SESAME in the case of external radiotherapy

  17. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  18. Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

    International Nuclear Information System (INIS)

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  19. Review of models applicable to accident aerosols

    International Nuclear Information System (INIS)

    Estimations of potential airborne-particle releases are essential in safety assessments of nuclear-fuel facilities. This report is a review of aerosol behavior models that have potential applications for predicting aerosol characteristics in compartments containing accident-generated aerosol sources. Such characterization of the accident-generated aerosols is a necessary step toward estimating their eventual release in any accident scenario. Existing aerosol models can predict the size distribution, concentration, and composition of aerosols as they are acted on by ventilation, diffusion, gravity, coagulation, and other phenomena. Models developed in the fields of fluid mechanics, indoor air pollution, and nuclear-reactor accidents are reviewed with this nuclear fuel facility application in mind. The various capabilities of modeling aerosol behavior are tabulated and discussed, and recommendations are made for applying the models to problems of differing complexity

  20. ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Chatelard, P., E-mail: patrick.chatelard@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France); Reinke, N.; Arndt, S. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50677 Köln (Germany); Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France); Eckel, J. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50677 Köln (Germany); Jacq, F.; Marchetto, C.; Mun, C.; Piar, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France)

    2014-06-01

    The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants.

  1. Algebraic geometric codes with applications

    Institute of Scientific and Technical Information of China (English)

    CHEN Hao

    2007-01-01

    The theory of linear error-correcting codes from algebraic geomet-ric curves (algebraic geometric (AG) codes or geometric Goppa codes) has been well-developed since the work of Goppa and Tsfasman, Vladut, and Zink in 1981-1982. In this paper we introduce to readers some recent progress in algebraic geometric codes and their applications in quantum error-correcting codes, secure multi-party computation and the construction of good binary codes.

  2. Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

    International Nuclear Information System (INIS)

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs

  3. Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; Jow, H-N (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA))

    1990-12-01

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs.

  4. Development status of Severe Accident Analysis Code SAMPSON

    Energy Technology Data Exchange (ETDEWEB)

    Iwashita, Tsuyoshi; Ujita, Hiroshi [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan)

    2000-11-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

  5. Artificial intelligence applications in accident management

    International Nuclear Information System (INIS)

    For nuclear power plant accident management, there are some addition concerns: linking AI systems to live data streams must be mastered; techniques for processing sensor inputs with varying data quality need to be provided; systems responsiveness to changing plant conditions and multiple user requests should, in general, be improved; there is a need for porting applications from specialized AI machines onto conventional computer hardware without incurring unacceptable performance penalties; human factors guidelines are required for new user interfaces in AI applications; methods for verification and validation of AI-based systems must be developed; and, finally, there is a need for proven methods to evaluate use effectiveness and firmly establish the benefits of AI-based accident management systems. (orig./GL)

  6. Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

  7. Validation of system codes for plant application on selected experiments

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

  8. Validation of system codes for plant application on selected experiments

    International Nuclear Information System (INIS)

    For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

  9. Recent SCDAP/RELAP5 code applications and improvements

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  10. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  11. Status of the GAMMA-FR code validation - TES pipe rupture accident of HCCR TBS

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung Gon; Lee, Dong Won; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of); Merrill, Brad J. [Idaho National Laboratory, Atomic (United States); Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    GAMMA-FR code to code validation is conducted and it shows reasonable agreement, however, near wall effect on the effective thermal conductivity needs to be investigated for better results. The GAMMA-FR code was scheduled for validation during the next two years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases. The GAMMA-FR (Gas Multicomponent Mixture Transient Analysis for Fusion Reactors) code is an in-house system analysis code to predict the thermal hydraulic and chemical reaction phenomena expected to occur during the thermo-fluid transients in a nuclear fusion system. A safety analysis of the Korea TBS (Test Blanket System) for ITER (International Thermonuclear Experimental Reactor) is underway using this code. This paper describes validation strategy of GAMMA-FR and current status of the validation study with respect to 'TES pipe rupture accident of ITER TBM'.

  12. Status of the GAMMA-FR code validation - TES pipe rupture accident of HCCR TBS

    International Nuclear Information System (INIS)

    GAMMA-FR code to code validation is conducted and it shows reasonable agreement, however, near wall effect on the effective thermal conductivity needs to be investigated for better results. The GAMMA-FR code was scheduled for validation during the next two years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases. The GAMMA-FR (Gas Multicomponent Mixture Transient Analysis for Fusion Reactors) code is an in-house system analysis code to predict the thermal hydraulic and chemical reaction phenomena expected to occur during the thermo-fluid transients in a nuclear fusion system. A safety analysis of the Korea TBS (Test Blanket System) for ITER (International Thermonuclear Experimental Reactor) is underway using this code. This paper describes validation strategy of GAMMA-FR and current status of the validation study with respect to 'TES pipe rupture accident of ITER TBM'

  13. Proceedings of the Seminar on Methods and Codes for Assessing the off-site consequences of nuclear accidents. Volume 1

    International Nuclear Information System (INIS)

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled 'methods for assessing the radiological impact of accidents' (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  14. User's manual of ART code for analyzing fission product transport behavior during core meltdown accident

    International Nuclear Information System (INIS)

    In a probabilistic risk assessment (PRA) it has been recognized that a core meltdown accident with a large amount of fission products released to the environment is a dominant contributor to public risk. For the evaluation of the risk, information about source terms are inevitable. In order to analyze fission product transport behavior and to evaluate source terms during a core meltdown accident, the ART code has been developed. The ART code has the following features: (1) It can treat fission product transport behavior both in a primary system and a containment system, (2) It models fission product transport caused by both gas flow and liquid flow, and (3) It includes a detailed model about transport behavior of aerosols which are released in quantity during a core meltdown accident. This report is a user's manual for the ART code and includes description of modeling, input/output data and a sample run. (author)

  15. ASTEC V2 severe accident integral code: Fission product modelling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

  16. Network Coding Fundamentals and Applications

    CERN Document Server

    Medard, Muriel

    2011-01-01

    Network coding is a field of information and coding theory and is a method of attaining maximum information flow in a network. This book is an ideal introduction for the communications and network engineer, working in research and development, who needs an intuitive introduction to network coding and to the increased performance and reliability it offers in many applications. This book is an ideal introduction for the research and development communications and network engineer who needs an intuitive introduction to the theory and wishes to understand the increased performance and reliabil

  17. Safety analysis of MNSR reactor during reactivity insertion accident using the validated code PARET

    International Nuclear Information System (INIS)

    In the framework of the IAEA CRP project (J7.10.10) on 'Safety significance of postulated initiating events for various types of research reactors and assessment of analytical tools' the Syrian team contributed in the assessment of computational codes related to the safety analysis of research reactors. During the project implementation the codes PARET and MERSAT have been tested, modified and verified regarding specific phenomena related to safety analysis of research reactors. In the framework of this contribution the code PARET has been applied to model the core of the Syrian MNSR reactor. The code analysis includes the simulation of steady state operation and a group of selected reactivity insertion accident (RIA) including the design basis accidents dealing with the insertion of total available excess reactivity

  18. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H{sub 2}/air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author).

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

    International Nuclear Information System (INIS)

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

  20. Radiological accident analysis with simulation codes; Analisis de accidentes radiologicos con codigos de simulacion

    Energy Technology Data Exchange (ETDEWEB)

    Brucker, R.; Munoz, A.; Rodriguez, J.

    2011-07-01

    The scope of radiological analysis is to calculate the dose received by the public and by an operator in the control room in case of an accident. Simulation software are needed for that kind of analysis in order to solve differential equations (radionuclides transport equations), to simulate the accident scenario, and to calculate the dose. This article presents the main radionuclide transport codes (several cases simulated with RADTRAD v3.03 are detailed), dose calculation programs, and atmospheric dispersion coefficients calculation software. (Author) 10 refs.

  1. Multi-step approach to Code-coupling for progression induced severe accidents in CANDU NPPs (MACPISA-CANDU)

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, D.J.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada); Giannotti, W.; D' Auria, F. [Univ. of Pisa, Dept. of Mechanical, Nuclear and Production Engineering, Pisa (Italy)

    2009-07-01

    This paper reviews the progression of severe accidents, describes computer codes currently employed for analysis of severe accidents and outlines a new methodology to modelling the progression of severe accidents in CANDU nuclear power plants (NPPs) called the Multi-step Approach to Code-coupling for Progression Induced Severe Accidents in CANDU NPPs (MACPISA-CANDU). The MACPISA-CANDU methodology was used to couple the U.S. NRC codes SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4) and MELCOR (1.8.5) in order to model a small break loss of coolant accident with loss of emergency coolant injection (SBLOCA-LOECI) under natural circulation in a CANDU 6 NPP. Using this model it was shown that the sheath temperature did not exceed the zirconium melting temperature of 2098 K and hence the progression of the severe accident was terminated as expected. (author)

  2. Use and development of coupled computer codes for the analysis of accidents at nuclear power plants. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Computer codes are widely used in Member States for the analysis of safety at nuclear power plants (NPPs). Coupling of computer codes, a further tool for safety analysis, is especially beneficial to safety analysis. The significantly increased capacity of new computation technology has made it possible to switch to a newer generation of computer codes, which are capable of representing physical phenomena in detail and include a more precise consideration of multidimensional effects. The coupling of advanced, best estimate computer codes is an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Coupling of computer codes is very advantageous for studies which relate to licensing of new NPPs, safety upgrading programmes for existing plants, periodic safety reviews, renewal of operating licences, use of safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, justification for lifetime extensions, development of new emergency operating procedures, analysis of operational events and development of accident management programmes. In this connection, the OECD/NEA Working Group on the Analysis and Management of Accidents (GAMA) recently highlighted the application of coupled computer codes as an area of 'high collective interest'. Coupled computer codes are being developed in many Member States independently or within small groups composed of several technical organizations. These developments revealed that there are many types and methods of code coupling. In this context, it was believed that an exchange of views and experience while addressing these problems at an international meeting could contribute to the more efficient and reliable use of advanced computer codes in nuclear safety applications. The present publication constitutes the report on the Technical Meeting on Progress in the Development and Use of Coupled Codes for Accident

  3. Development of a dose assessment computer code for the NPP severe accident

    International Nuclear Information System (INIS)

    A real-time emergency dose assessment computer code called KEDA (KAIST NPP Emergency Dose Assessment) has been developed for the NPP severe accident. A new mathematical model which can calculate cloud shine has been developed and implemented in the code. KEDA considers the specific Korean situations(complex topography, orientals' thyroid metabolism, continuous washout, etc.), and provides functions of dose-monitoring and automatic decision-making. To verify the code results, KEDA has been compared with an NRC officially certified code, RASCAL, for eight hypertical accident scenarios. Through the comparison, KEDA has been proved to provide reasonable results. Qualitative sensitivity analysis also the been performed for potentially important six input parameters, and the trends of the dose v.s. down-wind distance curve have been analyzed comparing with the physical phenomena occurred in the real atmosphere. The source term and meteorological conditions are turned out to be the most important input parameters. KEDA also has been applied to simulate Kori site and a hyperthetical accident with semi-real meteorological data has been simulated and analyzed

  4. Development of Parameter Network for Accident Management Applications

    Energy Technology Data Exchange (ETDEWEB)

    Pak, Sukyoung; Ahemd, Rizwan; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Jung Taek; Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    When a severe accident happens, it is hard to obtain the necessary information to understand of internal status because of the failure or damage of instrumentation and control systems. We learned the lessons from Fukushima accident that internal instrumentation system should be secured and must have ability to react in serious conditions. While there might be a number of methods to reinforce the integrity of instrumentation systems, we focused on the use of redundant behavior of plant parameters without additional hardware installation. Specifically, the objective of this study is to estimate the replaced value which is able to identify internal status by using set of available signals when it is impossible to use instrumentation information in a severe accident, which is the continuation of the paper which was submitted at the last KNS meeting. The concept of the VPN was suggested to improve the quality of parameters particularly to be logged during severe accidents in NPPs using a software based approach, and quantize the importance of each parameter for further maintenance. In the future, we will continue to perform the same analysis to other accident scenarios and extend the spectrum of initial conditions so that we are able to get more sets of VPNs and ANN models to predict the behavior of accident scenarios. The suggested method has the uncertainty underlain in the analysis code for severe accidents. However, In case of failure to the safety critical instrumentation, the information from the VPN would be available to carry out safety management operation.

  5. Simulation of rod ejection accident in a WWER-1000 Nuclear Reactor by using PARCS code

    International Nuclear Information System (INIS)

    Highlights: • REA in WWER-1000 Nuclear Reactor was simulated. • PARCS v2.7 and WIMSD-5B codes were used. • PARCS was validated for steady-state and transient processes. • Temperature reactivity coefficient was calculated. • TH block of PARCS v2.7 code was used. - Abstract: The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism housing that results in the complete ejection of a rod cluster control assembly from the reactor core. The consequences of the mechanical failure are a rapid positive reactivity insertion and an increase in the local power peaking with high local energy deposition in the fuel assembly, accompanied by an initial pressure increase in the reactor cooling system. In this study, the REA has been simulated in a WWER-1000 reactor by using WIMSD-5B and PARCS v2.7 codes. First, macroscopic cross-sections have been calculated for various types of fuel assemblies using WIMSD-5B. Results have been fed as input to PARCS v2.7 code. Steady-state, transient and specially thermal–hydraulic feedback blocks of PARCS code have been handled in this simulation. Finally, results have been compared with Final Safety Analysis Report of WWER-1000 reactor. The results show a great similarity and confirm the ability of PARCS code in simulation of transient accidents

  6. The coupling algorithm between fuel pin and coolant channel in the European Accident Code EAC-2

    International Nuclear Information System (INIS)

    In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release. This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules are also briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC. (orig.)

  7. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  8. SACO-1: a fast-running LMFBR accident-analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed.

  9. SHETEMP: a computer code for calculation of fuel temperature behavior under reactivity initiated accidents

    International Nuclear Information System (INIS)

    A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  11. A2 Code - Internal Accident Report. Does it ring a bell?

    CERN Multimedia

    HSE Unit

    2015-01-01

    A2 Code* - It is under this designation (used by the CERN community) that the form for internal accident reports is hidden. More specifically it refers to the CERN Safety Code A2 “Reporting of Accidents and Near Misses” (EDMS: 335502 or here via the official Safety Rules website).   Which events should be declared? All accidental events, which cause or could have caused injuries or damage to property or the environment, must be reported especially if they involve: a) a member of the personnel, visitor, temporary labourer or contractor if it occurred on the CERN site or between sites. b) a member of the personnel if it occurred while commuting or during duty travel. Who can fill in the report? The reporting of occurred accidents or near misses should be made by the person involved or by any direct or indirect witness of the event as soon as possible after the event. Contribute to the improvement of Safety within the Organizatio...

  12. MABEL-1. A code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    The MABEL-1 code has been written to investigate the deformation, of fuel pin cladding and its effects on fuel pin temperature transients during a loss-of-coolant accident. The code considers a single fuel pin with heated fuel concentric within the cladding. The fuel pin temperature distribution is evaluated using a one-dimensional conduction model with heat transfer to the coolant represented by an input set of heat transfer coefficients. The cladding deformation is calculated using the code CANSWEL, which assumes all strain to be elastic or creep and models the creep under a multi-axial stress system by a spring/dashpot combination undergoing alternate relaxation and elastic strain. (author)

  13. An Evaluation Methodology Development and Application Process for Severe Accident Safety Issue Resolution

    OpenAIRE

    Martin, Robert P.

    2012-01-01

    A general evaluation methodology development and application process (EMDAP) paradigm is described for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe-accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management...

  14. Simulation of THAI HD-12 test with the Accident Source Term Evaluation Code (ASTEC)

    Energy Technology Data Exchange (ETDEWEB)

    Braehler, Thimo; Koch, Marco K. [Bochum Univ. (DE). Chair of Energy Systems and Energy Economics (LEE)

    2010-05-15

    In case of a hypothetical severe accident in nuclear power plants, hydrogen can be generated and released into the containment. The generation of hydrogen during an accident is caused by an exothermal reaction from dissociated oxygen of water coolant in the primary circuit with the cladding material of the fuel rods. If an ignition of the released hydrogen occurs the integrity of the containment can be endangered, due to the pressure and temperature rise during the combustion process. The maximum pressure and temperature is influenced by geometric configurations and different initial conditions like hydrogen and steam concentration in the ambient of the containment. According to the generated hydrogen can lead to dry H2-concentration of between 17% and 20% for homogenous distributed atmosphere in a pressurized water reactor, if 100% of the fuel cladding material oxidizes. But depending on the accident scenario, the steam concentration varies in range of 20 to 70 %. Obstacles in the path of the flame front can increases the grade of turbulence, which among others, enhance the burning rate and accelerates the flame speed. Dry air mixtures with hydrogen concentrations over 4 vol.-% are ignitable. Due to buoyancy effects, the direction of the flame propagation has a distinct influence on the combustion process. The flammability limit of hydrogen air mixtures is shifted from 4 vol.-% for upward directed combustion to 8 vol.-% in downward direction. The THAI HD-test (Hydrogen Deflagration) series contains 29 experiments with different initial conditions like temperature, pressure, steam and hydrogen concentration in vertical up- and downward directed flame propagation. These had been carried out to investigate the phenomenology of hydrogen combustion and to provide the necessary data for the validation and modelling of computer codes. The basis for the code validation contains a broad number of experiments, but there are less data available for sufficient large scale

  15. Fuel Behavior Simulation Code FEMAXI-FBR Development for SFR Core Disruptive Accident Analysis

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been developing ASTERIA-FBR code system for SFR core disruptive accident analysis to contribute as a part of the regulation activity for Japanese prototype FBR, MONJU. The ASTERIA-FBR code system consists of detailed fuel behavior analysis module (FEMAXI-FBR), neutronic Monte-Carlo calculation module (GMVP), and thermal hydraulic module (CONCORD). The calculation scope of the ASTERIA-FBR covers the initiating, transitional and post disassembly expansion processes. The FEMAXI-FBR is based on LWR fuel behavior simulation code FEMAXI-6 and modified the material properties and the calculation models under steady state and transient operational condition. The FEMAXI-FBR has been verified in steady state calculations compared with those of SAS-4A code. Furthermore, the code has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient calculations. Through these verification and validation, good agreement has been obtained with the FP-gas release ratio, the fuel restructuring, the gap width between pellet and cladding, and the fuel pin failure position. (author)

  16. Computer code for the analyses of reactivity initiated accident of heavy water moderated and cooled research reactor 'EUREKA-2D'

    International Nuclear Information System (INIS)

    Codes, such as EUREKA and EUREKA-2 have been developed to analyze the reactivity initiated accident for light water reactor. These codes could not be applied directly for the analyses of heavy water moderated and cooled research reactor which are different from light water reactor not only on operation condition but also on reactor kinetic constants. EUREKA-2D which is modified EUREKA-2 is a code for the analyses of reactivity initiated accident of heavy water research reactors. Following items are modified: 1) reactor kinetic constants. 2) thermodynamic properties of coolant. 3) heat transfer equations. The feature of EUREKA-2D and an example of analysis are described in this report. (author)

  17. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  18. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    International Nuclear Information System (INIS)

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences

  19. A Mobile Application Prototype using Network Coding

    DEFF Research Database (Denmark)

    Pedersen, Morten Videbæk; Heide, Janus; Fitzek, Frank;

    2010-01-01

    This paper looks into implementation details of network coding for a mobile application running on commercial mobile phones. We describe the necessary coding operations and algorithms that implements them. The coding algorithms forms the basis for a implementation in C++ and Symbian C++. We report...

  20. Simulation of the core degradation phase of the Fukushima accidents using the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Bonneville, H., E-mail: herve.bonneville@irsn.fr; Luciani, A.

    2014-06-01

    The French Institute for Nuclear Safety and Radioprotection (IRSN) attempts to simulate the Fukushima accidents using the ASTEC integral code. This paper summarizes the main results of the simulations conducted before the beginning of the OECD/NEA/CSNI Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. The first analysis carried out concerned the unit 2 transient. Results were considered as satisfactory being quite consistent with measures reported by TEPCO and similar computations performed with MELCOR or MAAP. Knowledge gained from PWR practice and different lectures available in the open literature for BWR provided valuable technical elements to explain observations or to validate assumptions. Leakage model from the containment up to the refuelling bay through the head flange seal was very efficient to retrieve pressure evolution inside the dry well. Extension of the model to reactor number 3 gave also results quite consistent with what similar codes computed. However for both reactors some figures characteristic of the transient as hydrogen production are liable to vary a lot if models for bottom and top nozzles are added which has not been done in reference computation due to present lack of data. Uncertainties with simulation of accident on reactor number 1 are rather large due to the scarcity of data. Further, as the measurement points were quasi absent for most of the first 24 h there is no reference to compare to simulation results. Bottom vessel head failure is predicted but due to the high number of penetrations the mechanical failure models developed for PWR may not be so relevant for BWR.

  1. Development of a severe accident module of a nuclear power plant based in the MELCOR nuclear code and its incorporation to the room simulator

    International Nuclear Information System (INIS)

    This work describes the development of the Severe Accidents Module (MAS) based on the Code MELCOR and its incorporation to the Simulator of Classroom of the Group of Nuclear Engineering of the Engineering Faculty (GrINFI) of the National Autonomous University of Mexico (UNAM). The module of Severe Accidents has the purpose of counting with installed and operational capacity for the simulation of accident sequences with capacitation purposes, training, analysis and design. A shallow description of SimAula is presented, and the philosophy used to obtain the interactive version of MELCOR are discussed, as well as its implementation in the atmosphere of SimAula. Finally, after confirming the correct operation of the development of the tool, some possible topics are discussed for specific applications of the MAS. (Author)

  2. A Mobile Application Prototype using Network Coding

    OpenAIRE

    Pedersen, Morten Videbæk; Heide, Janus; Fitzek, Frank; Larsen, Torben

    2010-01-01

    This paper looks into implementation details of network coding for a mobile application running on commercial mobile phones. We describe the necessary coding operations and algorithms that implements them. The coding algorithms forms the basis for a implementation in C++ and Symbian C++. We report on practical measurement results of coding throughput and energy consumption for a single-source multiple-sinks network, with and without recoding at the sinks. These results confirm that network cod...

  3. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  4. Use and development of coupled computer codes for the analysis of accidents at nuclear power plants. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Computer codes are widely used in Member States for the analysis of safety at nuclear power plants (NPPs). Coupling of computer codes, a further tool for safety analysis, is especially beneficial to safety analysis. The significantly increased capacity of new computation technology has made it possible to switch to a newer generation of computer codes, which are capable of representing physical phenomena in detail and include a more precise consideration of multidimensional effects. The coupling of advanced, best estimate computer codes is an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Coupling of computer codes is very advantageous for studies which relate to licensing of new NPPs, safety upgrading programmes for existing plants, periodic safety reviews, renewal of operating licences, use of safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, justification for lifetime extensions, development of new emergency operating procedures, analysis of operational events and development of accident management programmes. In this connection, the OECD/NEA Working Group on the Analysis and Management of Accidents (GAMA) recently highlighted the application of coupled computer codes as an area of 'high collective interest'. Coupled computer codes are being developed in many Member States independently or within small groups composed of several technical organizations. These developments revealed that there are many types and methods of code coupling. In this context, it was believed that an exchange of views and experience while addressing these problems at an international meeting could contribute to the more efficient and reliable use of advanced computer codes in nuclear safety applications. The present publication constitutes the report on the Technical Meeting on Progress in the Development and Use of Coupled Codes for Accident

  5. C++ application development with Code::Blocks

    CERN Document Server

    Modak, Biplab Kumar

    2013-01-01

    This is a comprehensive tutorial with step-by-step instructions on how to develop applications with Code::Blocks.This book is for C++ developers who wish to use Code::Blocks to create applications with a consistent look and feel across multiple platforms. This book assumes that you are familiar with the basics of the C++ programming language.

  6. Development of Lower Plenum Molten Pool Module of Severe Accident Analysis Code in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Son, Donggun; Kim, Dong-Ha; Park, Rae-Jun; Bae, Jun-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shim, Suk-Ku; Marigomen, Ralph [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2014-10-15

    To simulate a severe accident progression of nuclear power plant and forecast reactor pressure vessel failure, we develop computational software called COMPASS (COre Meltdown Progression Accident Simulation Software) for whole physical phenomena inside the reactor pressure vessel from a core heat-up to a vessel failure. As a part of COMPASS project, in the first phase of COMPASS development (2011 - 2014), we focused on the molten pool behavior in the lower plenum, heat-up and ablation of reactor vessel wall. Input from the core module of COMPASS is relocated melt composition and mass in time. Molten pool behavior is described based on the lumped parameter model. Heat transfers in between oxidic, metallic molten pools, overlying water, steam and debris bed are considered in the present study. The models and correlations used in this study are appropriately selected by the physical conditions of severe accident progression. Interaction between molten pools and reactor vessel wall is also simulated based on the lumped parameter model. Heat transfers between oxidic pool, thin crust of oxidic pool and reactor vessel wall are considered and we solve simple energy balance equations for the crust thickness of oxidic pool and reactor vessel wall. As a result, we simulate a benchmark calculation for APR1400 nuclear power plant, with assumption of relocated mass from the core is constant in time such that 0.2ton/sec. We discuss about the molten pool behavior and wall ablation, to validate our models and correlations used in the COMPASS. Stand-alone SIMPLE program is developed as the lower plenum molten pool module for the COMPASS in-vessel severe accident analysis code. SIMPLE program formulates the mass and energy balance for water, steam, particulate debris bed, molten corium pools and oxidic crust from the first principle and uses models and correlations as the constitutive relations for the governing equations. Limited steam table and the material properties are provided

  7. OSSA - An optimized approach to severe accident management: EPR application

    International Nuclear Information System (INIS)

    There is a recognized need to provide nuclear power plant technical staff with structured guidance for response to a potential severe accident condition involving core damage and potential release of fission products to the environment. Over the past ten years, many plants worldwide have implemented such guidance for their emergency technical support center teams either by following one of the generic approaches, or by developing fully independent approaches. There are many lessons to be learned from the experience of the past decade, in developing, implementing, and validating severe accident management guidance. Also, though numerous basic approaches exist which share common principles, there are differences in the methodology and application of the guidelines. AREVA/Framatome-ANP is developing an optimized approach to severe accident management guidance in a project called OSSA ('Operating Strategies for Severe Accidents'). There are still numerous operating power plants which have yet to implement severe accident management programs. For these, the option to use an updated approach which makes full use of lessons learned and experience, is seen as a major advantage. Very few of the current approaches covers all operating plant states, including shutdown states with the primary system closed and open. Although it is not necessary to develop an entirely new approach in order to add this capability, the opportunity has been taken to develop revised full scope guidance covering all plant states in addition to the fuel in the fuel building. The EPR includes at the design phase systems and measures to minimize the risk of severe accident and to mitigate such potential scenarios. This presents a difference in comparison with existing plant, for which severe accidents where not considered in the design. Thought developed for all type of plants, OSSA will also be applied on the EPR, with adaptations designed to take into account its favourable situation in that field

  8. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    Energy Technology Data Exchange (ETDEWEB)

    Hermsmeyer, S. [European Commission JRC, Petten (Netherlands). Inst. for Energy and Transport; Herranz, L.E.; Iglesias, R. [CIEMAT, Madrid (Spain); and others

    2015-07-15

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  9. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    International Nuclear Information System (INIS)

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  10. Evidence from glycine transfer RNA of a frozen accident at the dawn of the genetic code

    Directory of Open Access Journals (Sweden)

    Tate Warren P

    2008-12-01

    Full Text Available Abstract Background Transfer RNA (tRNA is the means by which the cell translates DNA sequence into protein according to the rules of the genetic code. A credible proposition is that tRNA was formed from the duplication of an RNA hairpin half the length of the contemporary tRNA molecule, with the point at which the hairpins were joined marked by the canonical intron insertion position found today within tRNA genes. If these hairpins possessed a 3'-CCA terminus with different combinations of stem nucleotides (the ancestral operational RNA code, specific aminoacylation and perhaps participation in some form of noncoded protein synthesis might have occurred. However, the identity of the first tRNA and the initial steps in the origin of the genetic code remain elusive. Results Here we show evidence that glycine tRNA was the first tRNA, as revealed by a vestigial imprint in the anticodon loop sequences of contemporary descendents. This provides a plausible mechanism for the missing first step in the origin of the genetic code. In 448 of 466 glycine tRNA gene sequences from bacteria, archaea and eukaryote cytoplasm analyzed, CCA occurs immediately upstream of the canonical intron insertion position, suggesting the first anticodon (NCC for glycine has been captured from the 3'-terminal CCA of one of the interacting hairpins as a result of an ancestral ligation. Conclusion That this imprint (including the second and third nucleotides of the glycine tRNA anticodon has been retained through billions of years of evolution suggests Crick's 'frozen accident' hypothesis has validity for at least this very first step at the dawn of the genetic code. Reviewers This article was reviewed by Dr Eugene V. Koonin, Dr Rob Knight and Dr David H Ardell.

  11. International Code Assessment and Applications Program: Annual report

    International Nuclear Information System (INIS)

    This is the first annual report of the International Code Assessment and Applications Program (ICAP). The ICAP was organized by the Office of Nuclear Regulatory Research, United States Nuclear Regulatory Commission (USNRC) in 1985. The ICAP is an international cooperative reactor safety research program planned to continue over a period of approximately five years. To date, eleven European and Asian countries/organizations have joined the program through bilateral agreements with the USNRC. Seven proposed agreements are currently under negotiation. The primary mission of the ICAP is to provide independent assessment of the three major advanced computer codes (RELAP5, TRAC-PWR, and TRAC-BWR) developed by the USNRC. However, program activities can be expected to enhance the assessment process throughout member countries. The codes were developed to calculate the reactor plant response to transients and loss-of-coolant accidents. Accurate prediction of normal and abnormal plant response using the codes enhances procedures and regulations used for the safe operation of the plant and also provides technical basis for assessing the safety margin of future reactor plant designs. The ICAP is providing required assessment data that will contribute to quantification of the code uncertainty for each code. The first annual report is devoted to coverage of program activities and accomplishments during the period between April 1985 and March 1987

  12. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

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

    International Nuclear Information System (INIS)

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

  14. Hybrid codes: Methods and applications

    Energy Technology Data Exchange (ETDEWEB)

    Winske, D. (Los Alamos National Lab., NM (USA)); Omidi, N. (California Univ., San Diego, La Jolla, CA (USA))

    1991-01-01

    In this chapter we discuss hybrid'' algorithms used in the study of low frequency electromagnetic phenomena, where one or more ion species are treated kinetically via standard PIC methods used in particle codes and the electrons are treated as a single charge neutralizing massless fluid. Other types of hybrid models are possible, as discussed in Winske and Quest, but hybrid codes with particle ions and massless fluid electrons have become the most common for simulating space plasma physics phenomena in the last decade, as we discuss in this paper.

  15. Brief evaluation of the radiological hazards after a nuclear accident - description and mode of operation of this calculation code Orion

    International Nuclear Information System (INIS)

    The ORION code is designed to determine very quickly the immediate consequences (such as plume passage time, instantaneous maximum hazards irradiation, inhalation, deposit) due to an accident spreading out radioactive or chemical pollution into the atmosphere, from a source point, a stack release, (with heightening calculation) outspread sources (transport accident such as, for instance, road fire or car crash) or from a cylindrical cloud defined by different vertical sources (for instance pyrotechnical accident, missile firing...). The diffusion code DOURY type (french official methods) is written in FORTRAN. Data are entered in a conversational mode with auto-checking. Results are output to tables an isorisks curves drawn at map scales. At the Bruyeres-le-Chatel Radiation Protection Unit, a team is on permanent duty, can carry out results in a few minutes and transmit the evaluation by TELEFAX anywhere on the National territory

  16. Ruthenium release modelling in air under severe accident conditions using the MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Beuzet, E.; Lamy, J.S. [EDF R and D, 1 avenue du General de Gaulle, F-92140 Clamart (France); Perron, H. [EDF R and D, Avenue des Renardieres, Ecuelles, F-77818 Moret sur Loing (France); Simoni, E. [Institut de Physique Nucleaire, Universite de Paris Sud XI, F-91406 Orsay (France)

    2010-07-01

    In a nuclear power plant (NPP), in some situations of low probability of severe accidents, an air ingress into the vessel occurs. Air is a highly oxidizing atmosphere that can lead to an enhanced core degradation affecting the release of Fission Products (FPs) to the environment (source term). Indeed, Zircaloy-4 cladding oxidation by air yields 85% more heat than by steam. Besides, UO{sub 2} can be oxidised to UO{sub 2+x} and mixed with Zr, which may lead to a decrease of the fuel melting temperature. Finally, air atmosphere can enhance the FPs release, noticeably that of ruthenium. Ruthenium is of particular interest for two main reasons: first, its high radiotoxicity due to its short and long half-life isotopes ({sup 103}Ru and {sup 106}Ru respectively) and second, its ability to form highly volatile compounds such as ruthenium gaseous tetra-oxide (RuO{sub 4}). Considering that the oxygen affinity decreases between cladding, fuel and ruthenium inclusions, it is of great need to understand the phenomena governing fuel oxidation by air and ruthenium release as prerequisites for the source term issues. A review of existing data on ruthenium release, controlled by fuel oxidation, leads us to implement a new model in the EDF version of MAAP4 severe accident code (Modular Accident Analysis Program). This model takes into account the fuel stoichiometric deviation and the oxygen partial pressure evolution inside the fuel to simulate its oxidation by air. Ruthenium is then oxidised. Its oxides are released by volatilisation above the fuel. All the different ruthenium oxides formed and released are taken into consideration in the model, in terms of their particular reaction constants. In this way, partial pressures of ruthenium oxides are given in the atmosphere so that it is possible to know the fraction of ruthenium released in the atmosphere. This new model has been assessed against an analytical test of FPs release in air atmosphere performed at CEA (VERCORS RT8). The

  17. Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    1998-10-01

    Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

  18. An Evaluation Methodology Development and Application Process for Severe Accident Safety Issue Resolution

    Directory of Open Access Journals (Sweden)

    Robert P. Martin

    2012-01-01

    Full Text Available A general evaluation methodology development and application process (EMDAP paradigm is described for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe-accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management guidelines. The basic framework described in this paper extends the top-down, bottom-up strategy described in the U.S Nuclear Regulatory Commission Regulatory Guide 1.203 to severe accident evaluations addressing U.S. NRC expectation for plant design certification applications.

  19. An application of probabilistic safety assessment methods to model aircraft systems and accidents

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

    1998-08-01

    A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

  20. On the application of near accident data to risk analysis of major accidents

    International Nuclear Information System (INIS)

    Major accidents are low frequency high consequence events which are not well supported by conventional statistical methods due to data scarcity. In the absence or shortage of major accident direct data, the use of partially related data of near accidentsaccident precursor data – has drawn much attention. In the present work, a methodology has been proposed based on hierarchical Bayesian analysis and accident precursor data to risk analysis of major accidents. While hierarchical Bayesian analysis facilitates incorporation of generic data into the analysis, the dependency and interaction between accident and near accident data can be encoded via a multinomial likelihood function. We applied the proposed methodology to risk analysis of offshore blowouts and demonstrated its outperformance compared to conventional approaches. - Highlights: • Probabilistic risk analysis is applied to model major accidents. • Two-stage Bayesian updating is used to generate informative distributions. • Accident precursor data are used to develop likelihood function. • A multinomial likelihood function is introduced to model dependencies among data

  1. A first accident simulation for Angra-1 power plant using the ALMOD computer code

    International Nuclear Information System (INIS)

    The acquisition of the Almod computer code from GRS-Munich to CNEN has permited doing calculations of transients in PWR nuclear power plants, in which doesn't occur loss of coolant. The implementation of the german computer code Almod and its application in the calculation of Angra-1, a nuclear power plant different from the KWU power plants, demanded study and models adaptation; and due to economic reasons simplifications and optimizations were necessary. The first results define the analytical potential of the computer code, confirm the adequacy of the adaptations done and provide relevant conclusions about the Angra-1 safety analysis, showing at the same time areas in which the model can be applied or simply improved. (Author)

  2. Application of RS Codes in Decoding QR Code

    Institute of Scientific and Technical Information of China (English)

    Zhu Suxia(朱素霞); Ji Zhenzhou; Cao Zhiyan

    2003-01-01

    The QR Code is a 2-dimensional matrix code with high error correction capability. It employs RS codes to generate error correction codewords in encoding and recover errors and damages in decoding. This paper presents several QR Code's virtues, analyzes RS decoding algorithm and gives a software flow chart of decoding the QR Code with RS decoding algorithm.

  3. An Application of CICCT Accident Categories to Aviation Accidents in 1988-2004

    Science.gov (United States)

    Evans, Joni K.

    2007-01-01

    Interventions or technologies developed to improve aviation safety often focus on specific causes or accident categories. Evaluation of the potential effectiveness of those interventions is dependent upon mapping the historical aviation accidents into those same accident categories. To that end, the United States civil aviation accidents occurring between 1988 and 2004 (n=26,117) were assigned accident categories based upon the taxonomy developed by the CAST/ICAO Common Taxonomy Team (CICTT). Results are presented separately for four main categories of flight rules: Part 121 (large commercial air carriers), Scheduled Part 135 (commuter airlines), Non-Scheduled Part 135 (on-demand air taxi) and Part 91 (general aviation). Injuries and aircraft damage are summarized by year and by accident category.

  4. RSM modelling of an ATWS accident simulated by the ALMOD code: methodological and practical achievement

    International Nuclear Information System (INIS)

    A simulation study of a PWR station black-out ATWS has been performed by applying Response Surface Methodology (RSM) on the data obtained by inspecting the ALMOD code. The case under study has shown that the a priori information which alone could be inadequate, is optimally utilized if coupled with a preliminary sensitivity analysis through RSM techniques. In particular the engineering selection of the model variables and the rank order of the remaining ones had to be modified after an RSM preliminary sensitivity analysis. An other qualifying feature of the exercise is the use of randomization of the variables not included in the model in order to coherently exploit the methodology in its full efficiency. This procedure is able to give a figure of merit of the global importance of the neglected variables through the analysis of residuals. Results show that the proposed technique is an effective tool for selecting the most important accident variables and that the body of information gained is significant with respect to the number of observations performed

  5. Code Development on Aerosol Behavior under Severe Accident-Aerosol Coagulation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Ryu, Eun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The behaviors of the larger aerosol particles are described usually by continuum mechanics. The smallest particles have diameters less than the mean free path of gas phase molecules and the behavior of these particles can often be described well by free molecular physics. The vast majority of aerosol particles arising in reactor accident analyses have behaviors in the very complicated regime intermediate between the continuum mechanics and free molecular limit. The package includes initial inventories, release from fuel and debris, aerosol dynamics with vapor condensation and revaporization, deposition on structure surfaces, transport through flow paths, and removal by engineered safety features. Aerosol dynamic processes and the condensation and evaporation of fission product vapors after release from fuel are considered within each MELCOR control volume. The aerosol dynamics models are based on MAEROS, a multi-section, multicomponent aerosol dynamics code, but without calculation of condensation. Aerosols can deposit directly on surfaces such as heat structures and water pools, or can agglomerate and eventually fall out once they exceed the largest size specified by the user for the aerosol size distribution. Aerosols deposited on surfaces cannot currently be resuspended.

  6. Applicability of Phebus FP results to severe accident safety evaluations and management measures

    International Nuclear Information System (INIS)

    The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment

  7. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    International Nuclear Information System (INIS)

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

  8. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    Full Text Available The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN and uranium silicide (U3Si2. Candidate cladding materials include advanced stainless steel (FeCrAl and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2S-LWR, a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP Integrated Research Project (IRP is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S-LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I2S-LWR design

  9. Analysis of severe accident on OPR1000 PWR plant at low power and shutdown states with MAAP5 code

    International Nuclear Information System (INIS)

    The objective of this paper is to provide a brief description of severe accident analysis using computer codes in Korean OPR1000 Plant at low power and shutdown states. The results of the analysis are utilized in preparing the shutdown severe accident management guidelines (LPSD SAMG). As part of the efforts to prepare LPSD SAMG, analysis of severe accident is performed at low power and shutdown states with MAAP5 code. The Korean OPR1000 plant, a PWR plant with 2 hot legs and 4 cold legs is considered as a reference plant in the analysis. In this study, the scenarios are selected based on the plant operational states (POS) and dominant initiating events (IE) which cause the core damages. Typical scenarios are the loss of shutdown cooling (LSCS) at various primary coolant levels and stuck-opening of valves which prevent the low temperature over pressurization (LTOP) of primary system. As the analysis results, the core uncovery is expected in 2∼6 hours. The maximum temperature of core exit exceeds 649degC (SAMG entry temperature) in 3∼7 hours. The molten corium starts to relocate into lower head in 5∼13 hours and reactor vessel failure is occurred in 11∼14 hours. The above mentioned timings are utilized to choose the possible actions and the timing to implement those actions LPSD SAMG. Also based on the results, the environmental conditions that instruments may encounter in a severe accident are determined. (author)

  10. Accident analysis in the water loop of the nuclear engineering department of IPEN using the RELAP4 code

    International Nuclear Information System (INIS)

    A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting from (1) loss of coolant, (2) main pump failure and (3) power excursions, were studied. The computer code RELAP4/Mod.3 was employed as the principal tool of analysis. (Author)

  11. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety organization (JNES) is developing severe accident analysis codes in order to apply to the probabilistic safety assessment (PSA) for a typical fast breeder reactor (FBR). The AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary and the release fraction to the environment of fission products (FP). This report summarized results analyzed using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass (CVBP) scenario, and the containment failure scenario due to hydrogen deflagration or detonation. The results showed that the coolant temperature of the primary system and the secondary system in the PLOHS sequence increased at the almost same temperature, and the creep damage to the reactor coolant boundary became significant when coolant temperature exceeded about 1,100 K. The release fractions of FP in the CVBP case were estimated to be 0.99 for Xe, 0.14 for iodine, 0.44 for Cs and 0.01 for non-volatile tetravalent Ce. The release fractions of FP in the containment vessel failure case due to hydrogen burning were estimated to be 0.82 for Xe, 0.06 for iodine, 0.06 for Cs and 0.003 for non-volatile tetravalent Ce. In the present study, release fractions of FPs to the environment were obtained for the CVBP and the containment failure cases of the PLOHS accident sequence for the typical FBR plant. (author)

  12. Network Coding Protocols for Data Gathering Applications

    DEFF Research Database (Denmark)

    Nistor, Maricica; Roetter, Daniel Enrique Lucani; Barros, João

    2015-01-01

    Tunable sparse network coding (TSNC) with various sparsity levels of the coded packets and different feedback mechanisms is analysed in the context of data gathering applications in multi-hop networks. The goal is to minimize the completion time, i.e., the total time required to collect all data...... packets from the nodes while maintaining the per packet overhead at a minimum. We exploit two types of feedback, (1) the explicit feedback sent deliberately between nodes and (2) the implicit feedback emerged when a node hears its neighbour transmissions. Analytical bounds for a line network are derived...... using a fluid model, which is valid for any field size, various sparsity levels and the aforesaid feedback mechanisms. Our results show that implicit and explicit feedback mechanisms are instrumental in reducing the completion time for sparse codes....

  13. Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2012-01-01

    Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

  14. Writing robust C++ code for critical applications

    CERN Document Server

    CERN. Geneva

    2015-01-01

    **C++** is one of the most **complex**, expressive and powerful languages out there. However, its complexity makes it hard to write **robust** code. When using C++ to code **critical** applications, ensuring **reliability** is one of the key topics. Testing, debugging and profiling are all a major part of this kind of work. In the BE department we use C++ to write a big part of the controls system for beam operation, which implies putting a big focus on system stability and ensuring smooth operation. This talk will try to: - Highlight potential problems when writing C++ code, giving guidelines on writing defensive code that could have avoided such issues - Explain how to avoid common pitfalls (both in writing C++ code and at the debugging & profiling phase) - Showcase some tools and tricks useful to C++ development The attendees' proficiency in C++ should not be a concern. Anyone is free to join, even people that do not know C++, if only to learn the pitfalls a language may have. This may benefit f...

  15. Steady state and accident analysis of SCOR (simple compact reactor) with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Marie-Sophie Chenaud; Guy-Marie Gautier [CEA Cadarache- 13108 St Paul Lez Durance (France)

    2005-07-01

    Full text of publication follows: Within the framework of innovative reactors studies, the CEA was led to propose the SCOR design (Simple Compact Reactor). This design is based on a compact 600 MWe PWR and combines most of the advantages of innovative reactors. All main components such as the pressurizer, the canned pumps, the control rod mechanics and the dedicated heat exchangers on the passive residual heat removal system are integrated in the vessel.The only steam generator is located above the vessel in place of the upper head. The reactor operates at much lower primary circuit pressure than standard PWRs (85 bar instead of the usual 155 bar) and the power density is low (70 MW/m{sup 3} instead of 100 MW/m{sup 3} for the present PWRs). The reactivity being controlled by control rods and burnable poisons, there is no soluble boron. The elimination of a serious LOCA (Loss Of Coolant Accident) and the integrated residual heat removal system lead to enhanced safety with simple safety systems. Main features of the SCOR design and functional parameters have been previously reported. This paper focuses on the safety analysis of SCOR. Thermo hydraulic calculations have been run with the CATHARE code. Some calculations were run with the point kinetics module of CATHARE. Several transient simulations have been assessed. They concern a normal reactor trip from full power operation till refueling shutdown and accidental scenarios such as: - Loss of power, - Breaks from 0.02 m to 0.1 m on circuits connected to the vessel, - Steam generator tubes rupture, - Reactivity insertion by cold shock. Results of transient simulations enable us to conclude upon: - the increase of grace periods in comparison with standard PWRs if no safety systems operate besides emergency shutdown, - the expected efficiency of designed safety systems and in particular of the residual heat removal system in passive configuration even when integrated exchanger are dewatered. It will be retained that

  16. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    International Nuclear Information System (INIS)

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  17. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors

    International Nuclear Information System (INIS)

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  18. Uncertainty analysis with a view towards applications in accident consequence assessments

    International Nuclear Information System (INIS)

    Since the publication of the US-Reactor Safety Study WASH-1400 there has been an increasing interest to develop and apply methods which allow to quantify the uncertainty inherent in probabilistic risk assessments (PRAs) and accident consequence assessments (ACAs) for installations of the nuclear fuel cycle. Research and development in this area is forced by the fact that PRA and ACA are more and more used for comparative, decisive and fact finding studies initiated by industry and regulatory commissions. This report summarizes and reviews some of the main methods and gives some hints to do sensitivity and uncertainty analyses. Some first investigations aiming at the application of the method mentioned above to a submodel of the ACA-code UFOMOD (KfK) are presented. Sensitivity analyses and some uncertainty studies an important submodel of UFOMOD are carried out to identify the relevant parameters for subsequent uncertainty calculations. (orig./HP)

  19. Accident analysis of TEPCO's Fukushima Daiichi Nuclear Power Plant with the SAMPSON severe accident code. (2) Unit 1 analysis with improved debris relocation model

    International Nuclear Information System (INIS)

    On March 11, 2011, the Great Eastern Japan earthquake and the subsequent tsunami caused the station black out at TEPCO’s Fukushima Daiichi Nuclear Power Plants, and the events that followed led to core meltdowns. For assessment of the present core status, simulations have been performed with the SAMPSON severe accident code. The core debris relocation behaviors are newly investigated in this paper by applying the improved debris relocation model to the analysis of the Fukushima Daiichi unit 1 with SAMPSON code. The improvements to the model are as follows. (1) The velocity limiters and control rod guide tubes are newly taken into account. (2) The flow path of debris is modified so that it goes directly down to the lower plenum through the orifice, while in the old model, the debris had stayed on the core plate until the plate melted. In the plant analysis of unit 1 with the improved model, more than 96 wt% of the core debris is particulate. Much of debris, mainly composed of the fuel and zirconium particle, goes out of the core region through the orifice, while the debris falling on the velocity limiters is mainly composed of steel and control rod material particles. (author)

  20. Simulation Code Development and Its Applications

    Science.gov (United States)

    Li, Zenghai

    2015-10-01

    Under the support of the U.S. DOE SciDAC program, SLAC has been developing a suite of 3D parallel finite-element codes aimed at high-accuracy, high-fidelity electromagnetic and beam physics simulations for the design and optimization of next-generation particle accelerators. Running on the latest supercomputers, these codes have made great strides in advancing the state of the art in applied math and computer science at the petascale that enable the integrated modeling of electromagnetics, self-consistent Particle-In-Cell (PIC) particle dynamics as well as thermal, mechanical, and multi-physics effects. This paper will present the latest development and application of ACE3P to a wide range of accelerator projects.

  1. Thermo-physical properties of corium: development of an assessed data base for severe accident applications

    Energy Technology Data Exchange (ETDEWEB)

    Strizhov, V.F.; Galimov, R.G.; Ozrin, V.D. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Yu Zitserman, V.; Kobzev, G.I.; Fokin, L.R. [Institute of high temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Piluso, P. [CEA Cadarache (DEN/DTN/STRI), Lab. d' essais pour la Maitrise des Accidents graves, 13 - Saint Paul lez Durance (France); Chalaye, H. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)

    2007-07-01

    In a hypothetical case of a core melt-down scenarios a very high temperature would be reached (up to 3000 K). In this case, the materials of the core and structural materials (fuel, cladding, metallic alloys, concrete, etc.) could melt to form complex and aggressive mixtures called corium. Modelling of severe accident phenomena, code development and assessments of nuclear safety require a reliable knowledge of the thermophysical properties of corium at wide temperature range (below solidus temperature, between solidus and liquidus temperature and above the liquidus temperature). Common Russian-French project ISTC 3078, has been devoted to the development, assessment and recommendation for the establishment of a reliable thermophysical data base for severe accident applications. The project consists of two tasks related to properties of pure metallic (U, Zr, Fe, Cr, Ni) and oxide (UO{sub 2}, U{sub 3}O{sub 8}, U{sub 4}O{sub 9}, NiO, ZrO{sub 2}, Cr{sub 2}O{sub 3}, FeO, Fe{sub 2}O{sub 3}, Fe{sub 3}O{sub 4}, Al{sub 2}O{sub 3}, CaO, MgO, SiO{sub 2}, HfO{sub 2}, CeO{sub 2}) components, and mixtures relevant to severe accident conditions. Three categories of data (on UPAK classification) were considered: experimental data, critically evaluated data, and predicted data. The data of the first category is a result of specific experiment, data of the second category is a result of the analysis of data consistency and co-processing (expert and statistical) obtained in several experiments, data of the third category are based on model estimates, using correlations between different physical properties. The process of assessing, review and development of recommendation is described in the paper and illustrated by examples on thermophysical properties. (authors)

  2. Code comparison with MAAP 3.0 and March 3 (-STCP) for Nordic BWR and PWR plants to evaluate uncertainties in severe accident phenomena

    International Nuclear Information System (INIS)

    This study has been carried out within the framework of the Nordic NKA-AKTI-130-project whose participants are from Denmark, Finland and Sweden. The study is financed partly by the Nordic liaison committee for atomic energy and partly by national organisations. The goals of the study have been to achieve a common Nordic understanding of the capabilities of the severe accident codes MAAP 3.0 /1, 2/ and March 3-STCP /3/ and to evaluate uncertainties in severe accident phenomena by performing benchmark calculations and related sensitivity analyses for the existing Nordic power plants. The MAAP 3.0 code, which is an integrated thermal hydraulic and aerosol code, has been the main analysis tool in severe accident analyses in Sweden and Finland. Danish organisations have used the Source Term Code Package system (Mod 1.0) which is composed of several separate codes such as March 3, TRAPMELT etc. When plant specific design features are analyzed, a sensitivity type of study with a code system like MAAP 3.0 is an efficient tool. Experimental data for validation of code systems modelling the complex phenomena involved in severe accidents are, however, limited. It is in this situation valuable to compare models and results for two code systems developed by different organizations

  3. The primal application research of figure assimilation theory in the nuclear accident consequence forecast

    International Nuclear Information System (INIS)

    The deepgoing research of figure assimilation theory promotes many subjects' rapid development. This article outlooks the application of figure assimilation technique in the nuclear accident consequence forecast. The nuclear accident consequence forecast is a complicated system which needs rapidity and precision, so it is quiet difficult. but after the insertion of figure assimilation, it pushes on one step about the question. (authors)

  4. Detailed thermalhydraulic analysis of induced break severe accidents using the massively parallel CFD code TrioU/Priceles

    International Nuclear Information System (INIS)

    This paper reports the preliminary studies carried out with the CFD (computational fluid dynamics) code TrioU to study the natural gas circulation that may flow in the primary circuit of a pressurized water reactor during a high-pressure severe accident scenario. Two types of 3-dimensional simulations have been performed on one loop using a LES (large eddy simulations) approach. In the first type of calculations, the gas flow in the hot leg has been investigated with a simplified representation of the reactor vessel and the Steam Generator (SG) tubes. Structured and unstructured meshing have been tested on the full-scale geometry with and without radiative heat transfer modelling between walls and gas. The second type of calculations deals with the gas circulation in the SG. The first results show a good agreement with the available experimental data and provide some confidence in the TrioU code to simulate complex natural flows. (authors)

  5. Loss of Coolant Accident Analysis for 1MW PUSPATI Triga Mark II Research Reactor (RTP) Using MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Abd, Aziz Sadri [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    RTP is a pool type reactor cooled by natural circulation and the reactor core is located at the bottom of a demineralized water-filled aluminum liner tank of 2.0 meter diameter and 6.5 meter depth. The core assembly is composed of 100 cylindrical fuel rods including of 4 control rods in circular array. From the literature, development of thermal hydraulic analysis of RTP using computer code has not been well established. Therefore, establishment and development of appropriate thermal hydraulic safety analysis model is very critical to ensure the safety operation of the reactor. Hence, key thermal hydraulic parameters of RTP reactor operating under steady state and transient condition were investigated. In this paper, Loss Of Coolant Accident (LOCA) were calculated and analyzed and compared with corresponding values in Safety Analysis Report (SAR) 2008 and test report. PUSPATI Triga Mark II research reactor (RTP) has been operated at Malaysian Nuclear Agency since 1982 and primary cooling system was modified in 2010. Thermal hydraulic modeling of RTP of 1MWt has been successfully investigated with MARS-KS code. The calculated normal operation parameters have been compared with reactor Safety Analysis Report (SAR) and experimental data. Most of the thermal hydraulic parameters show good agreement with SAR and experimental data within an acceptable percentage error. The loss of coolant accident was simulated in case of leak of primary side heat exchanger gasket. The calculation result showed fast decrease of reactor pool level. About 5 minutes after the leak, reactor tank was fully depleted. Furthermore, claddings temperature was reached 1173.4K at 3270s which could result in failure of SS304 cladding. Based on the assessment, it is found that appropriate remedies including physical modifications or emergency procedures need be prepared to protect the reactor tank depletion by the heat exchanger leak accident.

  6. Loss of Coolant Accident Analysis for 1MW PUSPATI Triga Mark II Research Reactor (RTP) Using MARS-KS Code

    International Nuclear Information System (INIS)

    RTP is a pool type reactor cooled by natural circulation and the reactor core is located at the bottom of a demineralized water-filled aluminum liner tank of 2.0 meter diameter and 6.5 meter depth. The core assembly is composed of 100 cylindrical fuel rods including of 4 control rods in circular array. From the literature, development of thermal hydraulic analysis of RTP using computer code has not been well established. Therefore, establishment and development of appropriate thermal hydraulic safety analysis model is very critical to ensure the safety operation of the reactor. Hence, key thermal hydraulic parameters of RTP reactor operating under steady state and transient condition were investigated. In this paper, Loss Of Coolant Accident (LOCA) were calculated and analyzed and compared with corresponding values in Safety Analysis Report (SAR) 2008 and test report. PUSPATI Triga Mark II research reactor (RTP) has been operated at Malaysian Nuclear Agency since 1982 and primary cooling system was modified in 2010. Thermal hydraulic modeling of RTP of 1MWt has been successfully investigated with MARS-KS code. The calculated normal operation parameters have been compared with reactor Safety Analysis Report (SAR) and experimental data. Most of the thermal hydraulic parameters show good agreement with SAR and experimental data within an acceptable percentage error. The loss of coolant accident was simulated in case of leak of primary side heat exchanger gasket. The calculation result showed fast decrease of reactor pool level. About 5 minutes after the leak, reactor tank was fully depleted. Furthermore, claddings temperature was reached 1173.4K at 3270s which could result in failure of SS304 cladding. Based on the assessment, it is found that appropriate remedies including physical modifications or emergency procedures need be prepared to protect the reactor tank depletion by the heat exchanger leak accident

  7. MABEL 2: a code to analyse cladding deformation in a loss of coolant accident

    International Nuclear Information System (INIS)

    The calculation strategy of MABEL-2 and the hierarchy and purpose of its subroutines are described so that a programmer can readily identify both the overall structure of the code and the functions of its constituent parts. Also, to assist those who wish to examine the coding in detail, the common block variables are defined and a list is given of all variables used in the code, together with the subroutines in which they are used. (author)

  8. Two Applications of the Hamming-Golay Code

    Science.gov (United States)

    Liu, Andy

    2009-01-01

    In this paper, we give two unexpected applications of a Hamming code. The first one, also known as the "Hat Problem," is based on the fact that a small portion of the available code words are actually used in a Hamming code. The second one is a magic trick based on the fact that a Hamming code is perfect for single-error correction.

  9. MABEL 2: a code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    The code manual for MABEL-2 is written in four parts. Part 2 describes the equations programmed. The code is divided into a number of modules which are largely independent, namely the Geometry, Thermal-Hydraulic, Fuel and Cladding Temperature, Fuel Rod Internal Gas Pressure and Creep Modules. The equations in MABEL are described under these headings. (author)

  10. Study on virtual redundancy among process parameters for accident management applications

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Rizwan; Pak, Sukyoung; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Jungtaek; Park, Soo Yong; Ahn, Kwangil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The research at this point can be divided into three streams, focused on the development of self powered sensors and instrumentation, developing intelligent systems that can diagnose and accident type and developing indirect ways that is, methods to assess the safety critical parameters from other statistically related parameters. This first approach is quite expensive, second approach suffers from the limitation that infinite number of accident scenarios cannot be simulated. However, the only way to access the parameters during severe accidents is through simulation codes. Even-though, the process parameters data contain uncertainty, this is the only thing to start with severe accident management. International Nuclear Energy Research Initiative (Inert) project has started research to address various aspects of safety management during severe accidents. As a part of Inert team, we are investigating correlations among process parameters in such a way that safety critical information could be secured by means of other non-safety or virtual parameters during a severe accident. This is known as virtual redundancy of information. This will improve the availability of information in case one channel for information is lost. In this paper, we will discuss methodology, preliminary results and directions for further study. We found that several process parameters exhibit distinct variation pattern for a particular accident and several other parameters can also have the similar trends which strengthens the possibility of having virtual redundancy of information.

  11. Applications of bar code technology at nuclear power plants

    International Nuclear Information System (INIS)

    Bar code is an emerging technology that can eliminate handwritten and keyboard data-entry errors. With application-specific software, bar code technology can provide inventory control, reducing staff time and paperwork. This paper summarizes bar code technology, describes hardware commercially available, and reviews application software systems for use in nuclear power plants

  12. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  13. Applications of Derandomization Theory in Coding

    CERN Document Server

    Cheraghchi, Mahdi

    2011-01-01

    Randomized techniques play a fundamental role in theoretical computer science and discrete mathematics, in particular for the design of efficient algorithms and construction of combinatorial objects. The basic goal in derandomization theory is to eliminate or reduce the need for randomness in such randomized constructions. In this thesis, we explore some applications of the fundamental notions in derandomization theory to problems outside the core of theoretical computer science, and in particular, certain problems related to coding theory. First, we consider the wiretap channel problem which involves a communication system in which an intruder can eavesdrop a limited portion of the transmissions, and construct efficient and information-theoretically optimal communication protocols for this model. Then we consider the combinatorial group testing problem. In this classical problem, one aims to determine a set of defective items within a large population by asking a number of queries, where each query reveals w...

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

    International Nuclear Information System (INIS)

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

  15. Objective provision tree application to the effectiveness evaluation of accident management guidelines

    International Nuclear Information System (INIS)

    After the Fukushima accident in 2011, various lessons and safety enhancement action items were announced by national regulatory bodies. Among those items, the enforcement of procedural efficiency verification for accidents management guidelines including emergency operating procedures (EOPs), severe accident management guides (SAMGs) and extensive damage mitigating guidelines (EDMG) if applicable, was raised. The Objective Provision Tree (OPT) method is a top down approach which starts from the level of Defense in Depth (DiD), objectives and barriers, safety functions, challenges, mechanisms and finally ends with provisions. The benefit of OPT application to safety concerns includes that the OPT enables the comprehensive review for the verification of consistency and integrity of safety requirements for a specific safety issue. In this study, the preliminary framework for the application of OPT to the effectiveness evaluation of accident management guideline was introduced

  16. Applications of ASTEC integral code on a generic CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Gabriela, E-mail: gabriela.radu@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Prisecaru, Ilie [Power Engineering Department, University “Politehnica” of Bucharest, 313 Splaiul Independentei, Bucharest (Romania)

    2015-05-15

    Highlights: • Short overview of the models included in the ASTEC MCCI module. • MEDICIS/CPA coupled calculations for a generic CANDU6 reactor. • Two cases taking into account different pool/concrete interface models. - Abstract: In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon. The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core–concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction–correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust.

  17. Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

    Energy Technology Data Exchange (ETDEWEB)

    Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)

    1994-04-01

    Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions.

  18. An Analysis of Station Blackout Sequences Using MELCOR1.8.5 Code for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Song, Y. M.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    The Korea Atomic Energy Research Institute (KAERI) has been constructing severe accident analysis database (DB) under a National Nuclear R and D Program. Especially, MAAP (commercial code being widely used for industries) DB for many scenarios including station blackout (SBO) has been completed up to now. This report shows the analysis results for SBO scenarios using MELCOR code. These results will be used for the degree of completion after being compared with MAAP results. The developing strategy of MELCOR code is the same with that of MAAP DB. For the generation of data set, the Korean standard nuclear power plant (KSNP) has been selected as a reference plant and the eight SBO scenarios are chosen to be analyzed based on the PSA results (these eight scenarios accounted for 99 percent of occurrence frequency of total 197 SBO scenarios). Both thermal hydraulics (T/H) and source term analysis have been performed using MELCOR version 1.8.5 for the chosen scenarios. But only major T/H variables treated in the MAAP report are listed among the generated data set, which shows the characteristics of each scenario. These SBO results together with those of the other initiating events (to be analyzed in the future) will be used as inputs for DB construction and special value will be found in the comparing and complimentary process with MAAP DB

  19. A restructuring proposal based on MELCOR for severe accident analysis code development

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Hee; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    In order to develop a template based on existing MELCOR code, current data saving and transferring methods used in MELCOR are addressed first. Then a naming convention for the constructed module is suggested and an automatic program to convert old variables into new derived type variables has been developed. Finally, a restructured module for the SPR package has been developed to be applied to MELCOR. The current MELCOR code ensures a fixed-size storage for four different data types, and manages the variable-sized data within the storage limit by storing the data on the stacked packages. It uses pointer to identify the variables between the packages. This technique causes a difficult grasping of the meaning of the variables as well as memory waste. New features of FORTRAN90, however, make it possible to allocate the storage dynamically, and to use the user-defined data type which lead to a restructured module development for the SPR package. An efficient memory treatment and as easy understanding of the code are allowed in this developed module. The validation of the template has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. The template for the SPR package suggested in this report hints the extension of the template to the entire code. It is expected that the template will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models. 3 refs., 15 figs., 16 tabs. (Author)

  20. Analysis of Hydrogen Risk Mitigation System for Severe Accidents of EU-APR1400 Using MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mun Soo; Suh, Jung Soo; Bae, Byoung Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    According to the EUR (European Utility Requirements for LWR Nuclear Power Plants), it is mandatory that the HMS (Hydrogen Mitigation System) of the Eu-APR1400 should be equipped with a passive or automatic hydrogen control system. Considering this requirement, a PAR (Passive Autocatalytic Recombiner) system was adopted for the HMS of the Eu-APR1400. This passive HMS should be evaluated carefully in order to ensure that the HMS has adequate capacity to control hydrogen concentrations during severe accident conditions and to show that the system can satisfy the design requirements of the EUR. In this paper, analyses were carried out to examine the effectiveness of the HMS incorporated into the Eu- APR1400 design. These analyses were performed using the MAAP (Modular Accident Analysis Program) 4 code. in order to identify whether the HMS could control the average hydrogen concentrations in the containment, such that the concentration would not exceed 10 percent by volume: the analyses also considered whether there was the possibility of inadvertent hydrogen combustion in such processes as FA (Flame Acceleration) and DDT (Deflagration to Detonation Transition)

  1. Thermal-hydraulic system analysis using the MARS code for the transient steam generator tube rupture accident

    International Nuclear Information System (INIS)

    A postulated SGTR accident of the APR1400 was analysed using the best estimate safety analysis code, MARS. The main objective of this study is not only to provide physical insight into the system response of the APR1400 reactor during a SGTR but also to investigate the effect of reactor trip type of a HSGL and a LPP on the thermal-hydraulic system response. As for the tube rupture modelling method, double tube modelling was adopted. Broken U-tubes were modelled as a separate assembly of a single volume. The reactor trip type affected the overall progress of the major events. However, the effect on the thermal-hydraulic response of the plant was trivial. (author)

  2. Development and application of the waste code

    International Nuclear Information System (INIS)

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

  3. Coding Theory and Applications : 4th International Castle Meeting

    CERN Document Server

    Malonek, Paula; Vettori, Paolo

    2015-01-01

    The topics covered in this book, written by researchers at the forefront of their field, represent some of the most relevant research areas in modern coding theory: codes and combinatorial structures, algebraic geometric codes, group codes, quantum codes, convolutional codes, network coding and cryptography. The book includes a survey paper on the interconnections of coding theory with constrained systems, written by an invited speaker, as well as 37 cutting-edge research communications presented at the 4th International Castle Meeting on Coding Theory and Applications (4ICMCTA), held at the Castle of Palmela in September 2014. The event’s scientific program consisted of four invited talks and 39 regular talks by authors from 24 different countries. This conference provided an ideal opportunity for communicating new results, exchanging ideas, strengthening international cooperation, and introducing young researchers into the coding theory community.

  4. Simulation of a power pulse during loss of coolant accident in a CANDU-6 reactor by coupling the neutronic code PUMA and the thermalhydraulic code CATHENA

    International Nuclear Information System (INIS)

    In the frame of the safety analysis for a joint feasibility study (between Nucleoelectrica Argentina and Atomic Energy of Canada) of using slightly enriched uranium fuel (0.9 w% U235), Loss of Coolant Accidents (LOCAs) simulations were performed for Embalse NPP, a CANDU-6 type reactor (648. MWe gross). Being a reactor with a positive void reactivity coefficient, the void generation during the first seconds of LOCAs leads to an initial power increase, which is larger in the half of the reactor affected by the break. In order to simulate the power transient, which has a strong spatial variation in the flux and power distributions due to CANDU reactor features, two computer codes were used: the 3 dimensional diffusion, spatial kinetics neutronic program PUMA (developed in Argentina) and the thermal-hydraulics program CATHENA (developed in Atomic Energy of Canada). The codes were coupled by an iterative methodology: the CATHENA thermal-hydraulic simulation results (mainly temperatures of fuel and temperatures and densities of coolant) were used as input of the PUMA neutronic calculation, then the time dependent power distribution calculated by PUMA was applied as input for a new CATHENA calculation. The process was repeated up to convergence, which was obtained in a short number of iterations due to the relative minor effect of the power pulse and the strong influence of the break on the thermal-hydraulics Plant behavior during the analyzed time period. The method was utilized to simulate different accidental scenarios (break size and location, and initial conditions). (author)

  5. Development of severe accident Analysis Code SAMPSON in super simulator IMPACT' project

    Energy Technology Data Exchange (ETDEWEB)

    Morii, Tadashi; Ujita, Hiroshi; Vierow, Karen; Naitoh, Masanori [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan); Yamagishi, Makoto

    1999-07-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology', project Phase 1 have been completed. At the end of Phase 1, the Basic Single-, Two-, Multi-Phase Flow Analysis Modules of Various Coordinates have been parallelized. The physical models in the Boiling Transition Analysis Code and the Fluid-Structure Interaction Analysis Code have been completed and verified by comparison with basic experimental results. The verification study of the code was conducted in two steps. First, each analysis module was run independently and analysis results were compared against separate-effect experiment data. Verification analyses included: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex- Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. All comparison showed good agreement. Second, with the Simulation Supervisory Module, these analysis modules were executed concurrently in the parallel environment to demonstrate the code capability and integrity. (J.P.N.)

  6. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  7. Learning Vector Coding Methods of ART1 and Their Applications

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    As one of the unsupervised learning models, ART1 has been widely used in data mining or other fields, while coding of it's learning vector is very important. Their input vector coding methods and learning vector coding methods are described in detail. The corresponding applications are given.

  8. Integrated verification test of Severe Accident Analysis Code SAMPSON in super Simulation 'IMPACT' system

    Energy Technology Data Exchange (ETDEWEB)

    Ujita, Hiroshi; Naitoh, Masanori [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan); Karasawa, Hidetoshi; Miyagi, Kazumi

    1999-07-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology', project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analysed and phenomena occurred in scenarios can be simulated quantitatively reasonably considering the physical models used for the situation. (author)

  9. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    International Nuclear Information System (INIS)

    JNES is developing severe accident analysis codes in order to apply to the probability safety analysis (PSA) for a typical fast breeder reactor (FBR). AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary, and the discharge rate to the environment of fission products (FP). This report summarizes analysis results using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass scenario (CVBP) and the containment failure scenario by hydrogen deflagration or detonation. The coolant temperature of the primary system and the secondary system in the PLOHS sequence increases at the almost same temperature, and the creep damage to the reactor coolant boundary will become remarkable if coolant temperature exceeds about 1,100 K. In the CVBP scenario, when an intermediate heat exchanger is ruptured by creep and the boundary of the secondary system is failed, the path from the primary system to environment is formed. Then, the reactor vessel (RV) is failed and sodium in the primary coolant system releases into the reactor vessel room (RV room). Sodium of high temperature which fell in the RV room damages the floor liner, and generates hydrogen by a reaction with concrete. In addition the reactor core is exposed into atmosphere and the core temperature increases with decay heat and then volatile FP and non-volatile FP are released to the environment through the secondary system from the primary system. In the non-CVBP scenario which the intermediate heat exchanger does not fail by creep, core debris falls into the RV room after reactor vessel failure or evaporation of sodium coolant molten. FPs released from the reactor vessel are retained in the RV room, the primary system room, the containment dome and so on. The hydrogen generated by sodium-concrete reaction and

  10. Applications of Coding in Network Communications

    Science.gov (United States)

    Chang, Christopher SungWook

    2012-01-01

    This thesis uses the tool of network coding to investigate fast peer-to-peer file distribution, anonymous communication, robust network construction under uncertainty, and prioritized transmission. In a peer-to-peer file distribution system, we use a linear optimization approach to show that the network coding framework significantly simplifies…

  11. 'Turbo' coding for deep space applications

    DEFF Research Database (Denmark)

    Andersen, Jakob Dahl

    1995-01-01

    The performance of the `turbo' coding scheme is measured and an error floor is discovered. These residual errors are corrected with an outer BCH code. The complexity of the system is discussed, and for low data rates a realizable system operating at Eb/N0 below 0.2 dB is presented...

  12. Evidence from glycine transfer RNA of a frozen accident at the dawn of the genetic code

    OpenAIRE

    Tate Warren P; Bernhardt Harold S

    2008-01-01

    Abstract Background Transfer RNA (tRNA) is the means by which the cell translates DNA sequence into protein according to the rules of the genetic code. A credible proposition is that tRNA was formed from the duplication of an RNA hairpin half the length of the contemporary tRNA molecule, with the point at which the hairpins were joined marked by the canonical intron insertion position found today within tRNA genes. If these hairpins possessed a 3'-CCA terminus with different combinations of s...

  13. Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

    International Nuclear Information System (INIS)

    Highlights: ► We developed a new modelling of fuel oxidation and ruthenium release in the EDF version of the MAAP4 code. ► We validated this model against some VERCORS experiments. ► Ruthenium release prediction quantitatively and qualitatively well reproduced under air and steam atmospheres. - Abstract: In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103Ru and 106Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO4). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues. A review of existing data on these phenomena shows relatively good understanding. In terms of oxygen affinity, the fuel is oxidized before ruthenium, from UO2 to UO2+x. Its oxidation is a rate-controlling surface exchange reaction with the atmosphere, so that the stoichiometric deviation and oxygen partial pressure increase. High temperatures combined with the presence of

  14. Generalized polyphase representation and application to coding gain enhancement

    OpenAIRE

    Soman, Anand K.; Vaidyanathan, P.P.

    1994-01-01

    Generalized polyphase representations (GPP) have been mentioned in literature in the context of several applications. In this paper, we provide a characterization for what constitutes a valid GPP. Then, we study an application of GPP, namely in improving the coding gains of transform coding systems. We also prove several properties of the GPP.

  15. A POTENTIAL APPLICATION OF UNCERTAINTY ANALYSIS TO DOE-STD-3009-94 ACCIDENT ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Palmrose, D E; Yang, J M

    2007-05-10

    The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated accident releases with respect to the 25 rem Evaluation Guideline (EG) of DOE-STD-3009-94 CN3 (STD-3009). Historically, the classification of safety systems in a U.S. Department of Energy (DOE) nuclear facility's safety basis has involved how subject matter experts qualitatively view uncertainty in the STD-3009 Appendix A accident analysis methodology. Specifically, whether consequence uncertainty could be larger than previously evaluated so the site-specific accident consequences may challenge the EG. This paper assesses whether a potential uncertainty capability for MACCS2 could provide a stronger technical basis as to when the consequences from a design basis accident (DBA) truly challenges the 25 rem EG.

  16. Thailand Ranks Second in the World for Number of Road Accidents under Thailand’s Codes of Geometrical Design and Traffic Engineering Concept When Compared with AASHTO

    Directory of Open Access Journals (Sweden)

    CheewapattananuwongWeeradej

    2016-01-01

    Full Text Available Traffic problems in Bangkok have an influence on road users during peak hours. Especially, the traffic bottleneck on curves under the saturation flow situation must be remedied in order to increase the roadway capacity and speed. However, the appropriate speed for heavy vehicles is taken into consideration during off peak after the increasing lanes. This leads to the Rollover of heavy truck and rear-end collisions which are the main causes of vehicles accidents on curves. In addition, road accidents on curves account for the majority of all accidents in Thailand. According to the road accidents data collected in Thailand, 44 road deaths per 100,000 people, the country ranks second in the world for road accidents. When Thailand’s Code of Geometrical Design is compared with AASHTO (The American Association of State Highway and Transportation Officials, the super elevation length of Thailand’s Code is more than AASHTO. As a result, drivers are not made aware of the appropriate speed and the stooping sight distances (SSD on curves. Therefore, the Design of Traffic Signage under the Perception and Reaction Times (PRT for Thai Drivers will be taken into account.

  17. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104). [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kress, T. S. [comp.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.

  18. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

    International Nuclear Information System (INIS)

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

  19. Modelling of cladding oxidation by air under severe accident conditions with the MAAP 4 code

    International Nuclear Information System (INIS)

    In a nuclear power plant, air ingress into the vessel is a potential risk in some low probable situations of severe accidents. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of FP. This is particularly true speaking about ruthenium release, which can be significantly increased in the presence of air. This is a key issue due to the high radio-toxicity of ruthenium and its ability to form highly volatile oxides. The oxygen affinity is decreasing in priority from the Zircaloy cladding, to fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. As a first step, a phenomenological study has been carried out to characterize nitriding of the Zircaloy claddings. In summary, nitriding occurs preferentially when the oxygen has been consumed locally or in case of total oxygen starvation and when the cladding was slightly pre-oxidized. Just like oxidation, nitriding can be modeled in a simplified form as a cladding weight gain in terms of thickness. The model implemented in MAAP takes this into account as well as re-oxidation of the nitrides, in the case where oxygen is available again (especially during a reflood). Several correlations were thus integrated and a new one, called “KIT-EDF”, was developed, based on KIT separate-effect tests. The model has been implemented and validated against QUENCH-16 and QUENCH-10 experiments, studying the oxidation in air atmosphere of an assembly pre-oxidized in steam and finally quenched with water. The simulations give encouraging results since the modeling of nitriding effects has increased hydrogen production during reflood, as experimentally observed. The results of this study lead us to identify a number of perspectives for the future, namely taking into account the changes in the structure of the oxide layer during a

  20. From Barcode to QR Code Applications

    Directory of Open Access Journals (Sweden)

    László Várallyai

    2012-12-01

    Full Text Available This paper shows the Zsohár Horticulture Company in Nagyrákos, how they want to change their barcode identification system to QR code. They cultivate herbaceous, perpetual decorational plants, rock-garden, flower-bed and swamp perpetuals, decorational grasses and spices. A part of the perpetuals are evergreens, but most of them has special organs - such as onions, thick-, bulbous roots, "winter-proof" buds - so they can survive winter. In the first part of the paper I introduce the different barcode standards, how can it be printed and how can it be read. In the second part of the paper I give details about the quick response code (QR code and the two-dimensional (2D barcode. Third part of this paper illustrates the QR code usability in agriculture focused on the gardening.

  1. Zip Codes, Zip Codes for Limestone and Madison Counties, Published in 2014, Not Applicable scale, GIS.

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — This Zip Codes dataset, published at Not Applicable scale, was produced all or in part from Published Reports/Deeds information as of 2014. It is described as 'Zip...

  2. Analysis of the TMI-2 accident using ATHLET-CD

    International Nuclear Information System (INIS)

    One analyzed the simulation of the TMI-2 NPP accident making use of the ATHLET-CD code. One describes the accident sequence, the code structure and performs the comparative analysis of the calculated and the measured data. Simulation of thermohydraulic characteristics was a special success. Application of the codes promotes the NPP optimization, the reactor safety improvement and the risk reduction. The ATHLET-CD system ( the thermohydraulic analysis of leaks and transient processes at the reactor core disruption) will allow to evaluate the adequacy of the models included in the available codes to calculate severe accidents

  3. Bar code application to nuclear material accountancy

    International Nuclear Information System (INIS)

    For the purpose of efficient implementation of IAEA safeguards inspection, operators ought to prepare the information which is related to the strata for flow verification in a timely manner, such as physical inventory listing and summary of the fuel bundles. Today the use of bar code technique in tracing of products related data or counting number of items has been more and more applied to many facets of industry. From these points of view, the Japan Nuclear Fuel Company (NF) has been developing JNF Total Bar Code System. Now JNF has established an on-line input system of the fuel bundle accountability data by use of the bar code system to quickly prepare the information necessary for the inspection. As the first step, JNF implemented this bar code system at the flow verification to prepare physical inventory summary and location map of the fuel bundles in the storage. This paper reports that as a result of this, NF confirmed that this bar code system made it possible to input easily and quickly nuclear material accountancy information, and therefore this system is utilized as an effective and efficient measure of timely preparation for the inspection

  4. BAR-MOM Code and Its Application

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    BAR-MOM [1,2] code to calculate the height of the fission barrier Bf, the energy of the ground state, the compound nucleus stability by limit with respect to fission, i.e., the angular momentum(the spin value) Lmax at which the fission barrier disappears, the three principal axis moments of inertia at saddle point for a certain nucleus with atomic number Z, atomic mass number and angular momentum L for 19code to include the results for Z≥102[3] by using more recent parameterization of the Thomas Fermi fission

  5. Application of containment codes to LMFBRs in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.W.

    1977-01-01

    The application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs is described. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third proem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed.

  6. Optical Code Processing System, Device, and its Application

    Directory of Open Access Journals (Sweden)

    Naoya Wada

    2010-02-01

    Full Text Available Recent progress of optical code processing technology_ is explained. Ultra-high speed time domain, spectral domain, hybrid_ domain, and multiple optical code processing deices and systems are shown. As application of these technologies, OCDMA-PON, OPS network, and ultra high-speed optical clock generation will be demonstrated.

  7. Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Beuzet, Emilie, E-mail: emilie.beuzet@edf.f [EDF R and D, 1 Avenue du General de Gaulle, F-92140 Clamart (France); Lamy, Jean-Sylvestre, E-mail: jean-sylvestre.lamy@edf.f [EDF R and D, 1 Avenue du General de Gaulle, F-92140 Clamart (France); Bretault, Armelle, E-mail: armelle.bretault@edf.f [EDF R and D, 1 Avenue du General de Gaulle, F-92140 Clamart (France); Simoni, Eric, E-mail: simoni@ipno.in2p3.f [Institut de Physique Nucleaire, Universite Paris Sud XI, F-91406 Orsay (France)

    2011-04-15

    In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues. A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process. A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed

  8. Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

    International Nuclear Information System (INIS)

    In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues. A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process. A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-06-01

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

  11. Twenty years' application of agricultural countermeasures following the Chernobyl accident: lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Fesenko, S V [International Atomic Energy Agency, 1400 Vienna (Austria); Alexakhin, R M [Russian Institute of Agricultural Radiology and Agroecology, 249020 Obninsk (Russian Federation); Balonov, M I [International Atomic Energy Agency, 1400 Vienna (Austria); Bogdevich, I M [Research Institute for Soil Science and Agrochemistry, Minsk (Belarus); Howard, B J [Centre for Ecology and Hydrology, Lancaster Environment Centre, Library Avenue, Bailrigg, Lancaster LAI 4AP (United Kingdom); Kashparov, V A [Ukrainian Institute of Agricultural Radiology (UIAR), Mashinostroiteley Street 7, Chabany, Kiev Region 08162 (Ukraine); Sanzharova, N I [Russian Institute of Agricultural Radiology and Agroecology, 249020 Obninsk (Russian Federation); Panov, A V [Russian Institute of Agricultural Radiology and Agroecology, 249020 Obninsk (Russian Federation); Voigt, G [International Atomic Energy Agency, 1400 Vienna (Austria); Zhuchenka, Yu M [Research Institute of Radiology, 246000 Gomel (Belarus)

    2006-12-15

    The accident at the Chernobyl NPP (nuclear power plant) was the most serious ever to have occurred in the history of nuclear energy. The consumption of contaminated foodstuffs in affected areas was a significant source of irradiation for the population. A wide range of different countermeasures have been used to reduce exposure of people and to mitigate the consequences of the Chernobyl accident for agriculture in affected regions in Belarus, Russia and Ukraine. This paper for the first time summarises key data on countermeasure application over twenty years for all three countries and describes key lessons learnt from this experience. (review)

  12. Twenty years' application of agricultural countermeasures following the Chernobyl accident: lessons learned

    International Nuclear Information System (INIS)

    The accident at the Chernobyl NPP (nuclear power plant) was the most serious ever to have occurred in the history of nuclear energy. The consumption of contaminated foodstuffs in affected areas was a significant source of irradiation for the population. A wide range of different countermeasures have been used to reduce exposure of people and to mitigate the consequences of the Chernobyl accident for agriculture in affected regions in Belarus, Russia and Ukraine. This paper for the first time summarises key data on countermeasure application over twenty years for all three countries and describes key lessons learnt from this experience. (review)

  13. BAR-MOM code and its application

    International Nuclear Information System (INIS)

    BAR-MOM code for calculating the height of the fission barrier Bf , the energy of the ground state is presented; the compound nucleus stability by limit with respect to fission, i.e., the angular momentum (the spin value) Lmax at which the fission barrier disappears, the three principal axis moments of inertia at saddle point for a certain nucleus with atomic number Z, atomic mass number A and angular momentum L in units of ℎ for 19< Z<102, and the model used are introduced briefly. The generalized BAR-MOM code to include the results for Z ≥ 102 by using more recent parameterization of the Thomas Fermi fission barrier is also introduced briefly. We have learned the models used in Code BAR-MOM, and run it successfully and correctly for a certain nucleus with atomic mass number A, atomic number Z, and angular momentum L on PC by Fortran-90. The testing calculation values to check the implementation of the program show that the results of the present work are in good agreement with the original one

  14. Classifying hot water chemistry: Application of MULTIVARIATE STATISTICS - R code

    OpenAIRE

    Irawan, Dasapta Erwin; Gio, Prana Ugiana

    2016-01-01

    The following R code was used in this paper "Classifying hot water chemistry: Application of MULTIVARIATE STATISTICS" authors: Prihadi Sumintadireja1, Dasapta Erwin Irawan1, Yuano Rezky2, Prana Ugiana Gio3, Anggita Agustin1

  15. Two-phase computer codes for zero-gravity applications

    Energy Technology Data Exchange (ETDEWEB)

    Krotiuk, W.J.

    1986-10-01

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

  16. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  17. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

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

    International Nuclear Information System (INIS)

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

  19. TRIPOLI-4: Monte Carlo transport code functionalities and applications; TRIPOLI-4: code de transport Monte Carlo fonctionnalites et applications

    Energy Technology Data Exchange (ETDEWEB)

    Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), Service d' Etudes de Reacteurs et de Modelisation Avancee, 91 - Gif sur Yvette (France)

    2003-07-01

    Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

  20. Application of Inertia Ellipse in Code Marker Matching

    Institute of Scientific and Technical Information of China (English)

    XU Fang; JIANG Weiwei; HE Qing; HU Xiaobin

    2010-01-01

    In close-range photogrammetry, 3D information acquisition is based on image matching. The application of code marker helps to improve the level of automatic matching and the matching accuracy. This paper inyestigates the application of inertia ellipse algorithm to code marker matching. We can calculate the inertia ellipse of a target with a certain boundary. First, the method is applied to a single code marker; the angle and scaling are valid. Then, the paper introduces the multi code markers matching method by the inertia ellipse. Rotation and scaling changes of homonymy images can be calculated by inertia ellipse algorithm. These parameters can be used for code marker matching in arbitrary attitude close-range photogrammetry.

  1. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

    International Nuclear Information System (INIS)

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP)

  2. From Barcode to QR Code Applications

    OpenAIRE

    László Várallyai

    2012-01-01

    This paper shows the Zsohár Horticulture Company in Nagyrákos, how they want to change their barcode identification system to QR code. They cultivate herbaceous, perpetual decorational plants, rock-garden, flower-bed and swamp perpetuals, decorational grasses and spices. A part of the perpetuals are evergreens, but most of them has special organs - such as onions, thick-, bulbous roots, "winter-proof" buds - so they can survive winter. In the first part of the paper I introduce the different ...

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

    International Nuclear Information System (INIS)

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

  4. Application of Core Exit Temperature for Effective Safety Injection Strategy of Severe Accident Management Guidance

    International Nuclear Information System (INIS)

    Due to limited time for operator's action under the postulated severe accident, immediate and short term actions are needed and relevant strategies are constructed in the SAMG. Therefore, the SAMG includes a variety of information to assist the proper operator actions. Among these, pre-calculated graphs and formulas facilitate understanding of plant status and operator's action needed for accident mitigation. These are essential for ease of application and regarded as Computational Aids (CA). The representative example is the estimation of injection flow rates for removing decay heat and oxidation heat of core, and hydrogen generation rate, to mention a few. Most of all, calculation of the necessary injection flow rate is important in order to mitigate and/or terminate core damages. In estimating the flow rate for accident mitigation, Core Exit Temperature (CET) is utilized as a key variable. CET is considered most effective and reliable means for diagnosing core state. As such, CET has been adopted as a criterion transitioning from EOPs to SAMG. In this study, the necessary flow rate is calculated utilizing simple model with CET for RCS injection in mitigation strategy of SAMG. MELCOR simulation results are introduced for the calculation. A simple model of flow rate necessary for core heat removal is developed using CET data obtained from MELCOR simulations of OPR1000. The suggested model is expected to contribute on judging the core state in its coolability and required flow injection due to ease of application. More detailed analyses are needed to normalize by including additional accident scenarios

  5. Challenge Identification for the Objective Provision Tree Application to the Effectiveness Evaluation for the Accident Management Guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Kim, Hanchul; Lee, Sunghan [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    As a part of the OPT application for the effectiveness evaluation of the accident management guidelines, challenges which could threaten the safety functions required to maintain the safety, were identified. The identification of detailed provisions in terms of the accident management guidelines is being performed and the visualizing the identified elements of OPT is also under performance. With this logical structure of OPT, the provision of useful tool to evaluate the effectiveness of accident management guideline framework, is expected. The OPT method is a highly logical and top-down approach to identify the vulnerable aspect of the framework which includes the accident management guidelines, such as Emergency Operating Procedures (EOPs), Severe Accident Management Guides (SAMGs) and even Extensive Damage Mitigating Guidelines and FLEX guides. In virtue of this logical tool, the evaluation for the framework of the accident management guidelines was tried in this study.

  6. Coded optical time domain reflectometry: principle and applications

    Science.gov (United States)

    Park, Namkyoo; Lee, Jeonghwan; Park, Jonghan; Shim, Jae Gwang; Yoon, Hosung; Kim, Jin Hee; Kim, Kyoungmin; Byun, Jae-Oh; Bolognini, Gabriele; Lee, Duckey; Di Pasquale, Fabrizio

    2007-11-01

    In this paper, we will briefly outline our contributions for the physical realization of coded OTDR, along with its principles and also highlight recent key results related with its applications. For the communication network application, we report a multi-port / multi-wavelength, high-speed supervisory system for the in-service monitoring of a bidirectional WDM-PON system transmission line up to 16 ports x 32 nodes (512 users) capacity. Monitoring of individual branch traces up to 60 km was achieved with the application of a 127-bit simplex code, corresponding to a 7.5dB SNR coding gain effectively reducing the measurement time about 30 times when compared to conventional average mode OTDR. Transmission experiments showed negligible penalty from the monitoring system to the transmission signal quality, at a 2.5Gbps / 125Mbps (down / up stream) data rate. As an application to sensor network, a Raman scattering based coded-OTDR distributed temperature sensor system will be presented. Utilizing a 255-bit Simplex coded OTDR together with optimized sensing link (composed of cascaded fibers with different Raman coefficients), significant enhancement in the interrogation distance (19.5km from coding gain, and 9.6km from link-combination optimization) was achieved to result a total sensing range of 37km (at 17m/3K spatial/temperature resolution), employing a conventional off-shelf low power (80mW) laser diode.

  7. RAMSES-MHD: an AMR Godunov code for astrophysical applications

    Science.gov (United States)

    Fromang, S.; Hennebelle, P.; Teyssier, R.

    2005-12-01

    Godunov methods have proved in recent years to be very efficient numerical schemes to solve the hydrodynamic equations. Here, we present an extension of the 3D adaptative Mesh Refinament (AMR) code RAMSES (Teyssier 2002) to the equations of magnetohydrodynamics (MHD). The code uses the constrained transport scheme, which garantees that the divergence of the magnetic field is kept to zero to machine accuracy at all time. Different MHD Riemann solvers can be used, and the use of the MUSCL-Hancok approach combines a good accuracy with a fast exectution of the code. A variety of tests will illustrate the performances of the code and the possibilities offered by the AMR scheme. Future applications of the code are discussed.

  8. Chemical speciation code CHEMSPEC and its applications

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The adsorption and migration behavior of a radionuclide in geological media heavily depends on its chemical forms in a given chemical environment.In order to predict the temporal and spatial distribution of radionuclides around a disposal site when its canister is damaged,it is necessary to develop coupled chemical speciation-solute transport models and relevant software.For that reason,we wrote a new chemical speciation program CHEMSPEC.In this paper,the principles and structure of CHEMSPEC are briefly described,and the strategy and algorithms that were used in this code are interpreted in some detail,such as the measures adopted to prevent divergence in iteratively solving the mass balance equations,the "predictor-corrector" algorithm for calculation of the number and quantities of solid species formed,and the alternate use of "freezing" and "defreezing" oxidation states in handling of co-existent redox and precipitation equilibria.Four examples are given to illustrate CHEMSPEC’s features and capabilities.

  9. Application of numerical models and codes

    OpenAIRE

    Vyzikas, Thomas

    2014-01-01

    This report indicates the importance of numerical modelling in the modelling process, gradually builds the essential background theory in the fields of fluid mechanics, wave mechanics and numerical modelling, discusses a list of commonly used software and finally recommends which models are more suitable for different engineering applications in a marine renewable energy project.

  10. Parallelizing the MARS15 Code with MPI for shielding applications

    International Nuclear Information System (INIS)

    The MARS15 Monte Carlo code capabilities to deal with time-consuming deep penetration shielding problems and other computationally tough tasks in accelerator, detector and shielding applications, have been enhanced by a parallel processing option. It has been developed, implemented and tested on the Fermilab Accelerator Division Linux cluster and network of Sun workstations. The code uses MPI. It is scalable and demonstrates good performance. The general architecture of the code, specific uses of message passing, and effects of a scheduling on the performance and fault tolerance are described

  11. A study on PHWR moderator and severe accident analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Rhee, B. W.; Kim, D. H. [KAERI, Daejeon (Korea, Republic of); and others

    2012-04-15

    For the purpose of establishment of PHWR moderator and severe accident analysis system, the following works are performed. The main thermal-hydraulic phenomena are investigated and scaling analysis of the scaled down test facility design and fabrication are done to determine the scaling ratio based on the scaling law and practical constraints of the test facility. Theoretical background of the commercial CFD codes has been found out and their applicability and application conditions for the moderator circulation analysis are reviewed to develop the computer code requirement for the moderator 3-D analysis codes. Satisfactory analysis results against the STERN Lab. experiment showed the applicability of OpenFOAM and CUPID codes to moderator circulation analysis. For the development of various accident scenarios for establishing the DB for severe accident phenomena/progression, the level 1 and the level 2 PSA analysis results for Wolsong Unit 1 are reviewed and the most probable accident scenarios from the PDS event trees are selected. The latest ISAAC 4.03 version is used to predict the basic accident progression and the improvement items for the most up-to-date severe accident analysis issues analyzing function are derived. A basic system for the PHWR severe accident management decision making support system, SAMEX-CR is set up and requirement for the DB management system, SARDB-CR is derived to develop the implementation methodology for severe accident analysis DB management system.

  12. Stepwise integral scaling method and its application to severe accident phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M.; Zhang, G. [Purdue Univ., West Lafayette, IN (United States). School of Nuclear Engineering; No, H.C. [Korea Advanced Inst. of Science and Technology, Seoul (Korea, Republic of)

    1993-10-01

    Severe accidents in light water reactors are characterized by an occurrence of multiphase flow with complicated phase changes, chemical reaction and various bifurcation phenomena. Because of the inherent difficulties associated with full-scale testing, scaled down and simulation experiments are essential part of the severe accident analyses. However, one of the most significant shortcomings in the area is the lack of well-established and reliable scaling method and scaling criteria. In view of this, the stepwise integral scaling method is developed for severe accident analyses. This new scaling method is quite different from the conventional approach. However, its focus on dominant transport mechanisms and use of the integral response of the system make this method relatively simple to apply to very complicated multi-phase flow problems. In order to demonstrate its applicability and usefulness, three case studies have been made. The phenomena considered are (1) corium dispersion in DCH, (2) corium spreading in BWR MARK-I containment, and (3) incore boil-off and heating process. The results of these studies clearly indicate the effectiveness of their stepwise integral scaling method. Such a simple and systematic scaling method has not been previously available to severe accident analyses.

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

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

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

  14. Smart phone Application using Morse Code and Inaudible Frequency

    OpenAIRE

    Poonam Y. Pawar; Dimple S. Bhansali; Sandhya R. Borate; Poonam V. Deokate

    2013-01-01

    In this paper, the wireless communication using Morse code and inaudible frequency has been discussed. The application of this project is to transfer the limited information with the help of inaudible frequency and AAC. It is developed for Android smart phone. If user ever in a survival situation where phone service is not an option, can still use your phone to communicate over long distances with this application. The function of this application is to help the large number of users while ro...

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

    International Nuclear Information System (INIS)

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

  16. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  17. MELCOR analysis of the TMI-2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Boucheron, E.A.

    1990-01-01

    This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic predictions of severe accident phenomena and the radiological source team. The analysis of the TMI-2 standard problem allowed for comparison of the model predictions in MELCOR to plant data and to the results of more mechanistic analyses. This exercise was, therefore valuable for verifying and assessing the models in the code. The major trends in the TMI-2 accident are reasonably well predicted with MELCOR, even with its simplified modeling. Comparison of the calculated and measured results is presented and, based on this comparison, conclusions can be drawn concerning the applicability of MELCOR to severe accident analysis. 5 refs., 10 figs., 3 tabs.

  18. German offsite accident consequence model for nuclear facilities: further development and application

    International Nuclear Information System (INIS)

    The German Offsite Accident Consequence Model - first applied in the German Risk Study for nuclear power plants with light water reactors - has been further developed with the improvement of several important submodels in the areas of atmospheric dispersion, shielding effects of houses, and the foodchains. To aid interpretation, the presentation of results has been extended with special emphasis on the presentation of the loss of life expectancy. The accident consequence model has been further developed for application to risk assessments for other nuclear facilities, e.g., the liquid metal fast breeder reactor (SNR-300) and the high temperature gas cooled reactor. Moreover the model have been further developed in the area of optimal countermeasure strategies (sheltering, evacuation, etc.) in the case of the Central European conditions. Preliminary considerations has been performed in connection with safety goals on the basis of doses

  19. On application of CFD codes to problems of nuclear reactor safety

    International Nuclear Information System (INIS)

    The 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in May 2002 at Aix-en-Province, France, recommended formation of writing groups to report the need of guidelines for use and assessment of CFD in single-phase nuclear reactor safety problems, and on recommended extensions to CFD codes to meet the needs of two-phase problems in nuclear reactor safety. This recommendations was supported also by Working Group on the Analysis and Management of Accidents and led to formation oaf three Writing Groups. The first writing Group prepared a summary of existing best practice guidelines for single phase CFD analysis and made a recommendation on the need for nuclear reactor safety specific guidelines. The second Writing Group selected those nuclear reactor safety applications for which understanding requires or is significantly enhanced by single-phase CFD analysis, and proposed a methodology for establishing assesment matrices relevant to nuclear reactor safety applications. The third writing group performed a classification of nuclear reactor safety problems where extension of CFD to two-phase flow may bring real benefit, a classification of different modeling approaches, and specification and analysis of needs in terms of physical and numerical assessments. This presentation provides a review of these activities with the most important conclusions and recommendations (Authors)

  20. Review of the chronic exposure pathways models in MACCS [MELCOR Accident Consequence Code System] and several other well-known probabilistic risk assessment models

    International Nuclear Information System (INIS)

    The purpose of this report is to document the results of the work performed by the author in connection with the following task, performed for US Nuclear Regulatory Commission, (USNRC) Office of Nuclear Regulatory Research, Division of Systems Research: MACCS Chronic Exposure Pathway Models: Review the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and compare those models to the chronic exposure pathway models implemented in similar codes developed in countries that are members of the OECD. The chronic exposures concerned are via: the terrestrial food pathways, the water pathways, the long-term groundshine pathway, and the inhalation of resuspended radionuclides pathway. The USNRC has indicated during discussions of the task that the major effort should be spent on the terrestrial food pathways. There is one chapter for each of the categories of chronic exposure pathways listed above

  1. ASTEC V2.0 reactor applications on French PWR 900 MWe accident sequences and comparison with MAAP4

    Energy Technology Data Exchange (ETDEWEB)

    Lombard, Virginie; Azarian, Garo; Ducousso, Erik; Gandrille, Pascal, E-mail: pascal.gandrille@areva.com

    2014-06-01

    In the frame of the SARNET Severe Accident Network of Excellence an important task of partners is the assessment of the ASTEC integral code, considered today as the European reference code for evaluation of the source term. A code-to-code comparison between ASTEC V2.0 rev1 and MAAP 4.0.7 code versions has been performed by AREVA NP SAS on a French PWR 900 MWe. Two transients have been analyzed, focussing on in-vessel phenomena: total loss of feedwater (H2 sequence in the French nomenclature) and total loss of onsite and offsite power (H3 sequence). The detailed analysis shows an overall good agreement between both code results on thermal-hydraulics, hydrogen production and core degradation phenomena.

  2. Optical code division multiple access fundamentals and applications

    CERN Document Server

    Prucnal, Paul R

    2005-01-01

    Code-division multiple access (CDMA) technology has been widely adopted in cell phones. Its astonishing success has led many to evaluate the promise of this technology for optical networks. This field has come to be known as Optical CDMA (OCDMA). Surveying the field from its infancy to the current state, Optical Code Division Multiple Access: Fundamentals and Applications offers the first comprehensive treatment of OCDMA from technology to systems.The book opens with a historical perspective, demonstrating the growth and development of the technologies that would eventually evolve into today's

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

    International Nuclear Information System (INIS)

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

  4. Application of Structural Equations Modeling to assess relationship among Emotional Intelligence, General Health and Occupational Accidents

    Directory of Open Access Journals (Sweden)

    MOAMMAD KHANDAN

    2015-09-01

    Full Text Available ORIGINAL ARTICLEEmotional intelligence (EI has been subject of significant amounts of literature over the past two decades. However, little has been contributed to how emotional intelligence may be practically applied to enhance both accident prevention program and general health in workplaces. Purpose of this paper is to survey relationship among these variables in working society of Iran in 2014. As well as identify practical approaches to application of emotional intelligence skills to manage work change process.This was a cross-sectional study, conducted among all workers in functional units of a manufacturing company (n=178, located in a central province in Iran. Emotional intelligence assessed using Bradberry and Greaves’ questionnaire and Goldberg’s General Health Questionnaire (GHQ was the other tool used in the study. Descriptive statistics used to describe data by SPSS V22. Also, relationship among studied factors analyzed applying structural equations (SEM modeling by EQS software. Majority of workers (99.32% were male. Mean (SD age was 39.13 (8.23 also 64.19% of participants were married. Mean and standard deviation of EI score calculated 90.64 and 19.33, respectively. Also, results indicated that mean of GH score was 22.24 (±9.83. Analyzing relation between main variables (EI & GH with occupational accidents depicted that both of them are in significant relationship with accident (P<0.05. Regard to relationship between emotional intelligence and general health with occurrence of accidents in workplace and GH improvement with increase in EI, three strategies are recommended: appropriate job selection, Suitable training and intervention for workplace condition improvement.

  5. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  6. Development of a severe accident module of a nuclear power plant based in the MELCOR nuclear code and its incorporation to the room simulator; Desarrollo del modulo de accidentes severos de una central nucleoelectrica basado en el codigo nuclear MELCOR y su incorporacion al simulador de aula

    Energy Technology Data Exchange (ETDEWEB)

    Cortes M, F.S.; Ramos P, J.C.; Nelson E, P.; Chavez M, C. [Facultad de Ingenieria, Division de Ingenieria Electrica, Grupo de Ingenieria Nuclear, UNAM, Ciudad Universitaria, Distrito Federal (Mexico)]. E-mail: samuelcortes@correo.unam.mx

    2004-07-01

    This work describes the development of the Severe Accidents Module (MAS) based on the Code MELCOR and its incorporation to the Simulator of Classroom of the Group of Nuclear Engineering of the Engineering Faculty (GrINFI) of the National Autonomous University of Mexico (UNAM). The module of Severe Accidents has the purpose of counting with installed and operational capacity for the simulation of accident sequences with capacitation purposes, training, analysis and design. A shallow description of SimAula is presented, and the philosophy used to obtain the interactive version of MELCOR are discussed, as well as its implementation in the atmosphere of SimAula. Finally, after confirming the correct operation of the development of the tool, some possible topics are discussed for specific applications of the MAS. (Author)

  7. Development and first application of a new tool for the simulation of the initiating phase of a severe accident on SFR

    Science.gov (United States)

    Guyot, M.; Gubernatis, P.; Suteau, C.

    2014-06-01

    In order to improve the safety level of Sodium Fast Reactors, low probability events such as Hypothetical Core Disruptive Accident (HCDA) are analyzed for their potential consequences. The initiating phase of such accidents is of particular interest both for the prevention and the mitigation of routes leading to a large core disruption and recriticalities. Up to now, analysis of the initiating phase of HCDA has been performed with the SAS4A code. The SAS4A accident calculations are based on a multiple-channel approach, which requires that subassemblies or groups of similar subassemblies be represented together as independent channels. The SAS4A severe accident calculation scheme resorts to a simplified treatment in which an average pin is used to represent a channel. A point kinetics model coupled with a feedback reactivity model is also used to provide an estimate of the reactor power level. Both to increase the accuracy and decrease the uncertainties in the prediction of reactor safety margins, a new computational tool is currently under development at CEA Cadarache. The main features of this tool are the ability to provide a detailed sub-channel meshing of the sub-assembly as well as three-dimensional kinetics during severe accident conditions. To fulfill these goals, the fluid-dynamics SIMMER-III code has been coupled to the SNATCH solver using a MPI environment. This coupling allows both to compute the multi-phase and multi-component flows encountered in severe accident conditions and to model the power shape variation during voiding and melting of the different reactor materials. This new calculation scheme relies on a SAS-like multiple-channel treatment, where channel-to-channel heat and momentum exchanges are neglected. In this paper, an overview of the SIMMER-III/SNATCH coupled tool capabilities is provided. A first application of this new tool is also performed and compared with a SAS4A reference calculation. The new SIMMER-III/SNATCH tool proved to be

  8. List Decoding of Matrix-Product Codes from nested codes: an application to Quasi-Cyclic codes

    DEFF Research Database (Denmark)

    Hernando, Fernando; Høholdt, Tom; Ruano, Diego

    2012-01-01

    A list decoding algorithm for matrix-product codes is provided when $C_1,..., C_s$ are nested linear codes and $A$ is a non-singular by columns matrix. We estimate the probability of getting more than one codeword as output when the constituent codes are Reed-Solomon codes. We extend this list...... decoding algorithm for matrix-product codes with polynomial units, which are quasi-cyclic codes. Furthermore, it allows us to consider unique decoding for matrix-product codes with polynomial units....

  9. Decision making framework for application of forest countermeasures in the long term after the Chernobyl accident

    International Nuclear Information System (INIS)

    After the ChNPP accident a very large part of the territories covered by natural and artificial forests are contaminated with long-lived radionuclides, especially 137Cs. To protect people against exposure associated with forest contamination in the most affected regions of the NIS countries, countermeasures have been developed and recommended for the forest management. The paper presents a decision making framework to optimise forest countermeasures in the long term after the ChNPP accident. The approach presented is based on the analysis of the main exposure pathways and application of radiological, socio-economical and ecological criteria for the selection of optimal countermeasures strategies. Because of the diversity of these criteria modern decision support technologies based on multi-attributive analysis were applied. The results of the application of this approach are presented in a selected study area (Novozybkov district, Bryansk region, Russian Federation). The results prove and emphasize the need for a flexible technique to provide the optimised forest countermeasures taking into account radioecological, social and economic features of contaminated forests

  10. Quest for the real-time for the safety analysis code Cathare 2 used in the post-accident simulator Sipa

    Energy Technology Data Exchange (ETDEWEB)

    Ruby, A.; Antoni, O. [CEA Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Creach, V.; Dufeil, Ph. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Rose, Ch.; Iffenecker, F. [Electricite de France, 75 - Paris (France)

    2003-07-01

    The aim of the SCAR project is to use the CATHARE French thermal-hydraulic accident code in the SIPA simulator (Post-Accident Simulator) and extend SIPA to reactor cold shutdown states. The quest for real-time has been one of the key themes of the project since it began in 1997. The required CPU time depends on the computing power and on the ability of CATHARE to converge as fast as possible on the solution. Three main tasks have been scheduled to contain the lag between the simulation and the real-time: -1) Parallelism in CATHARE has been developed with shared-memory model (using OPEN MP). Standardized and adapted to the numerical method and the structure of CATHARE, it has enabled parallel tasks in 95% of the code with efficient parallel loops on the elements, and an optimized but limited parallelism in the solver. Validation has been carried out all along the task, ensuring the binary identity of results for 10 representative accident transients, whatever the number of processors used on each computer of the SCAR project. -2) Convergence has been improved for 20 CATHARE transients, ranging from the 100% full power state to cold-shutdown for maintenance state. A method based on the definition of maximum lag criteria in function of an estimated power of computers has been developed, revealing coding errors and leading to numerical improvements without any regression of physical law validation. A second phase has started in 2003 on another series of 25 transients within the simulator. -3) A techno-watch policy (using benchmarking) has allowed to keep up to date with progress in computer power throughout the duration of the project. It has consisted in comparing the performance of computers for 12 standard CATHARE input decks using an elementary time relevant of the computing machines for a given modeling of plant series. Furthermore, development validation and performance assessment tools have been developed at the same time. As a result of these three tasks

  11. Quest for the real-time for the safety analysis code Cathare 2 used in the post-accident simulator Sipa

    International Nuclear Information System (INIS)

    The aim of the SCAR project is to use the CATHARE French thermal-hydraulic accident code in the SIPA simulator (Post-Accident Simulator) and extend SIPA to reactor cold shutdown states. The quest for real-time has been one of the key themes of the project since it began in 1997. The required CPU time depends on the computing power and on the ability of CATHARE to converge as fast as possible on the solution. Three main tasks have been scheduled to contain the lag between the simulation and the real-time: -1) Parallelism in CATHARE has been developed with shared-memory model (using OPEN MP). Standardized and adapted to the numerical method and the structure of CATHARE, it has enabled parallel tasks in 95% of the code with efficient parallel loops on the elements, and an optimized but limited parallelism in the solver. Validation has been carried out all along the task, ensuring the binary identity of results for 10 representative accident transients, whatever the number of processors used on each computer of the SCAR project. -2) Convergence has been improved for 20 CATHARE transients, ranging from the 100% full power state to cold-shutdown for maintenance state. A method based on the definition of maximum lag criteria in function of an estimated power of computers has been developed, revealing coding errors and leading to numerical improvements without any regression of physical law validation. A second phase has started in 2003 on another series of 25 transients within the simulator. -3) A techno-watch policy (using benchmarking) has allowed to keep up to date with progress in computer power throughout the duration of the project. It has consisted in comparing the performance of computers for 12 standard CATHARE input decks using an elementary time relevant of the computing machines for a given modeling of plant series. Furthermore, development validation and performance assessment tools have been developed at the same time. As a result of these three tasks

  12. CONSUL code package application for LMFR core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

    2008-07-01

    CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

  13. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC

  14. Extracting recurrent scenarios from narrative texts using a Bayesian network: application to serious occupational accidents with movement disturbance.

    Science.gov (United States)

    Abdat, F; Leclercq, S; Cuny, X; Tissot, C

    2014-09-01

    A probabilistic approach has been developed to extract recurrent serious Occupational Accident with Movement Disturbance (OAMD) scenarios from narrative texts within a prevention framework. Relevant data extracted from 143 accounts was initially coded as logical combinations of generic accident factors. A Bayesian Network (BN)-based model was then built for OAMDs using these data and expert knowledge. A data clustering process was subsequently performed to group the OAMDs into similar classes from generic factor occurrence and pattern standpoints. Finally, the Most Probable Explanation (MPE) was evaluated and identified as the associated recurrent scenario for each class. Using this approach, 8 scenarios were extracted to describe 143 OAMDs in the construction and metallurgy sectors. Their recurrent nature is discussed. Probable generic factor combinations provide a fair representation of particularly serious OAMDs, as described in narrative texts. This work represents a real contribution to raising company awareness of the variety of circumstances, in which these accidents occur, to progressing in the prevention of such accidents and to developing an analysis framework dedicated to this kind of accident.

  15. The fuzzy set theory application to the analysis of accident progression event trees with phenomenological uncertainty issues

    International Nuclear Information System (INIS)

    Fuzzy set theory provides a formal framework for dealing with the imprecision and vagueness inherent in the expert judgement, and therefore it can be used for more effective analysis of accident progression of PRA where experts opinion is a major means for quantifying some event probabilities and uncertainties. In this paper, an example application of the fuzzy set theory is first made to a simple portion of a given accident progression event tree with typical qualitative fuzzy input data, and thereby computational algorithms suitable for application of the fuzzy set theory to the accident progression event tree analysis are identified and illustrated with example applications. Then the procedure used in the simple example is extended to extremely complex accident progression event trees with a number of phenomenological uncertainty issues, i.e., a typical plant damage state 'SEC' of the Zion Nuclear Power Plant risk assessment. The results show that the fuzzy averages of the fuzzy outcomes are very close to the mean values obtained by current methods. The main purpose of this paper is to provide a formal procedure for application of the fuzzy set theory to accident progression event trees with imprecise and qualitative branch probabilities and/or with a number of phenomenological uncertainty issues. (author)

  16. Development of a three-dimensional CDA analysis code. SIMMER-IV, and its first application to reactor case

    International Nuclear Information System (INIS)

    For the transition phase analysis of core disruptive accidents, the development of a three-dimensional reactor safety analysis code, SIMMER-IV, has been carried out based on the technology of the two-dimensional SIMMER-III code. The world first application of SIMMER-IV to a small-sized sodium-cooled fast reactor has also been attempted to clarify event progression in the early stage of the transition phase. This SIMMER-IV calculation is compared to the two-dimensional case calculated by SIMMER-III, neglecting the presence of control rod guide tubes. The present analysis with the three-dimensional representation suggests that the conventional scenario leading to rather early high-mobility fuel-pool formation is unrealistic and the degraded core tends to keep low mobility in the early stage of transition phase. (author)

  17. Tool Support for Inspecting the Code Quality of HPC Applications

    Energy Technology Data Exchange (ETDEWEB)

    Panas, T; Quinlan, D; Vuduc, R

    2007-03-16

    The nature of HPC application development encourages ad hoc design and implementation, rather than formal requirements analysis and design specification as is typical in software engineering. However, we cannot simply expect HPC developers to adopt formal software engineering processes wholesale, even while there is a need to improve software structure and quality to ensure future maintainability. Therefore, we propose tools that HPC developers can use at their discretion to obtain feedback on the structure and quality of their codes. This feedback would come in the form of code quality metrics and analyses, presented when necessary in intuitive and interactive visualizations. This paper summarizes our implementation of just such a tool, which we apply to a standard HPC benchmark as ''proof-of-concept.''

  18. Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

    International Nuclear Information System (INIS)

    A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes

  19. Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A. [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)]. E-mail: ahainoun@aec.org.sy; Alissa, S. [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)

    2005-01-01

    A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.

  20. Road Traffic Accident Analysis of Ajmer City Using Remote Sensing and GIS Technology

    Science.gov (United States)

    Bhalla, P.; Tripathi, S.; Palria, S.

    2014-12-01

    With advancement in technology, new and sophisticated models of vehicle are available and their numbers are increasing day by day. A traffic accident has multi-facet characteristics associated with it. In India 93% of crashes occur due to Human induced factor (wholly or partly). For proper traffic accident analysis use of GIS technology has become an inevitable tool. The traditional accident database is a summary spreadsheet format using codes and mileposts to denote location, type and severity of accidents. Geo-referenced accident database is location-referenced. It incorporates a GIS graphical interface with the accident information to allow for query searches on various accident attributes. Ajmer city, headquarter of Ajmer district, Rajasthan has been selected as the study area. According to Police records, 1531 accidents occur during 2009-2013. Maximum accident occurs in 2009 and the maximum death in 2013. Cars, jeeps, auto, pickup and tempo are mostly responsible for accidents and that the occurrence of accidents is mostly concentrated between 4PM to 10PM. GIS has proved to be a good tool for analyzing multifaceted nature of accidents. While road safety is a critical issue, yet it is handled in an adhoc manner. This Study is a demonstration of application of GIS for developing an efficient database on road accidents taking Ajmer City as a study. If such type of database is developed for other cities, a proper analysis of accidents can be undertaken and suitable management strategies for traffic regulation can be successfully proposed.

  1. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  2. Application of probabilistic safety assessment in CPR1000 severe accident prevention and mitigation analysis

    International Nuclear Information System (INIS)

    The relationship between probabilistic safety assessment (PSA) and severe accident study was discussed. Also how to apply PSA in severe accident prevention and mitigation was elaborated. PSA can find the plant vulnerabilities of severe accidents prevention and mitigation. Some modifications or improvements focusing on these vulnerabilities can be put forward. PSA also can assess the efficient of these actions for decision-making. According to CPR1000 unit severe accident analysis, an example for the process and method on how to use PSA to enhance the ability to deal with severe accident prevention and mitigation was set forth. (authors)

  3. Application of probabilistic safety analysis (PSA) approach to structuring accident mitigation systems of a PWR

    International Nuclear Information System (INIS)

    The safety evaluation technology of PWRs has already been improved substantially because of large-scale safety verification tests and improvement of accuracy in analyses. However, for structuring accident mitigation systems (AMS), the selection of appropriate systems from various AMS candidates mainly depends on engineering judgements by design engineers. So systematic designing process should be established. Reliability of each AMS forms the basis for reliability of safety plant design as a whole. Therefore, explicitly understanding characteristics of each AMS's reliability is very important for safety design. Based on these facts as a background, the limitation of improving reliability by strengthening redundancy of AMS mainly consisting of active components was clarified by applying PSA. At the same time, reliability and other characteristics of AMS mainly consisting of passive components were also clarified with PSA. Through these studies, it is proved that the application of PSA for structuring AMS is effective. (author)

  4. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000 degree F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion (''bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled

  5. Operator Design Methodology and Application in H.264 Entropy Coding

    Directory of Open Access Journals (Sweden)

    Ziyi Hu

    2010-11-01

    Full Text Available Currently ASIC applications, such as multimedia processing, require shorter time-to-market and lower cost of Non Recurring Engineering (NRE. Also, with the IC manufacturing technology developing continually, from transistor level to logic gate level, the size of design cells in digital circuits is increasing correspondingly. New design methodology is in urgent need to meet the requirement for the developing processing technology and shorter time-to-market in IC industry. This paper proposed the concepts and principles of operator design methodology, then focused on the entropy coding application based on the operators and finally presented the implementation results. The results show that with the proposed methodology, a comparable hardware performance can be obtained against the traditional standard cell based design flow. Furthermore, the design speed can be improved efficiently.

  6. Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity

  7. AXAIR: A Computer Code for SAR Assessment of Plume-Exposure Doses from Potential Process-Accident Releases to Atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Pillinger, W.L.

    2001-05-17

    This report describes the AXAIR computer code which is available to terminal users for evaluating the doses to man from exposure to the atmospheric plume from postulated stack or building-vent releases at the Savannah River Plant. The emphasis herein is on documentation of the methodology only. The total-body doses evaluated are those that would be exceeded only 0.5 percent of the time based on worst-sector, worst-case meteorological probability analysis. The associated doses to other body organs are given in the dose breakdowns by radionuclide, body organ and pathway.

  8. Network Coding Applications and Implementations on Mobile Devices

    DEFF Research Database (Denmark)

    Fitzek, Frank; Pedersen, Morten Videbæk; Heide, Janus;

    2010-01-01

    Network coding has attracted a lot of attention lately. The goal of this paper is to demonstrate that the implementation of network coding is feasible on mobile platforms. The paper will guide the reader through some examples and demonstrate uses for network coding. Furthermore the paper will also...... show that the implementation of network coding is feasible today on commercial mobile platforms....

  9. Development and application of traffic accident density estimation models using kernel density estimation

    OpenAIRE

    Seiji Hashimoto; Syuji Yoshiki; Ryoko Saeki; Yasuhiro Mimura; Ryosuke Ando; Shutaro Nanba

    2016-01-01

    Traffic accident frequency has been decreasing in Japan in recent years. Nevertheless, many accidents still occur on residential roads. Area-wide traffic calming measures including Zone 30, which discourages traffic by setting a speed limit of 30 km/h in residential areas, have been implemented. However, no objective implementation method has been established. Development of a model for traffic accident density estimation explained by GIS data can enable the determination of dangerous areas o...

  10. Quantitative information measurement and application for machine component classification codes

    Institute of Scientific and Technical Information of China (English)

    LI Ling-Feng; TAN Jian-rong; LIU Bo

    2005-01-01

    Information embodied in machine component classification codes has internal relation with the probability distribution of the code symbol. This paper presents a model considering codes as information source based on Shannon's information theory. Using information entropy, it preserves the mathematical form and quantitatively measures the information amount of a symbol and a bit in the machine component classification coding system. It also gets the maximum value of information amount and the corresponding coding scheme when the category of symbols is fixed. Samples are given to show how to evaluate the information amount of component codes and how to optimize a coding system.

  11. Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

    International Nuclear Information System (INIS)

    Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

  12. The Application Programming Interface for the PVMEXEC Program and Associated Code Coupling System

    Energy Technology Data Exchange (ETDEWEB)

    Walter L. Weaver III

    2005-03-01

    This report describes the Application Programming Interface for the PVMEXEC program and the code coupling systems that it implements. The information in the report is intended for programmers wanting to add a new code into the coupling system.

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

    International Nuclear Information System (INIS)

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

  14. UNICOS CPC6: Automated Code Generation for Process Control Applications

    CERN Document Server

    Fernandez Adiego, B; Prieto Barreiro, I

    2011-01-01

    The Continuous Process Control package (CPC) is one of the components of the CERN Unified Industrial Control System framework (UNICOS) [1]. As a part of this framework, UNICOS-CPC provides a well defined library of device types, amethodology and a set of tools to design and implement industrial control applications. The new CPC version uses the software factory UNICOS Application Builder (UAB) [2] to develop CPC applications. The CPC component is composed of several platform oriented plugins PLCs and SCADA) describing the structure and the format of the generated code. It uses a resource package where both, the library of device types and the generated file syntax, are defined. The UAB core is the generic part of this software, it discovers and calls dynamically the different plug-ins and provides the required common services. In this paper the UNICOS CPC6 package is introduced. It is composed of several plug-ins: the Instance generator and the Logic generator for both, Siemens and Schneider PLCs, the SCADA g...

  15. Utilisation of gas pipelines - Application of new codes

    Energy Technology Data Exchange (ETDEWEB)

    Bjoernsen, T. [Norske Veritas Industri Norge A/S, Hoevik (Norway)

    1997-12-31

    Current design codes are based upon requirements and safety philosophies introduced many decades ago. Few updates have been done compared to code development in other industries. The changes in the pipeline industry with new pipeline scenarios, standardisation and requirements to cost reduction have forced the industry to reconsider the current codes and look for improvements. Topics in this paper cover: Historical background on codes and standards; pipeline failure statistics; motivation for changes in current codes; limit state based design and safety, risk and reliability; status and standardisation and code development; discussion. 5 figs.

  16. The PHITS code for space applications: status and recent developments

    Science.gov (United States)

    Sihver, Lembit; Ploc, Ondrej; Sato, Tatsuhiko; Niita, Koji; Hashimoto, Shintaro; El-Jaby, Samy

    Since COSPAR 2012, the Particle and Heavy Ion Transport code System, PHITS, has been upgraded and released to the public [1]. The code has been improved and so has the contents of its package, such as the attached data libraries. In the new version, the intra-nuclear cascade models INCL4.6 and INC-ELF have been implemented as well as the Kurotama model for the total reaction cross sections. The accuracies of the new reaction models for transporting the galactic cosmic-rays were investigated by comparing with experimental data. The incorporation of these models has improved the capabilities of PHITS to perform particle transport simulations for different space applications. A methodology for assessing the pre-mission exposure of space crew aboard the ISS has been developed in terms of an effective dose equivalent [2]. PHITS was used to calculate the particle transport of the GCR and trapped radiation through the hull of the ISS. By using the predicted spectra, and fluence-to-dose conversion factors, the semi-empirical ISSCREM [3,4,5] code was then scaled to predict the effective dose equivalent. This methodology provides an opportunity for pre-flight predictions of the effective dose equivalent, which can be compared to post-flight estimates, and therefore offers a means to assess the impact of radiation exposure on ISS flight crew. We have also simulated [6] the protective curtain experiment, which was performed to test the efficiency of water-soaked hygienic tissue wipes and towels as a simple and cost-effective additional spacecraft shielding. The dose from the trapped particles and low energetic GCR, was significantly reduced, which shows that the protective curtains are efficient when they are applied on spacecraft at LEO. The results of these benchmark calculations, as well as the mentioned applications of PHITS to space dosimetry, will be presented. [1] T. Sato et al. J. Nucl. Sci. Technol. 50, 913-923 (2013). [2] S. El-Jaby, et al. Adv. Space Res. doi: http

  17. Application of bulk material commodity code in nuclear engineering

    International Nuclear Information System (INIS)

    The text details the signification and current status and difficulty of commodity code in the nuclear power engineering. By the applying condition of Ling Ao Phrase 2 Nuclear Power Plant there are several ways to create commodity code. Detail how to make commodity code structure and commodity code rule. And define material style, commodity code prefix, size and thickness etc. Then create commodity code. The other way is by user define to create commodity code. Next register specification in VPRM, make size range, thickness and branch fitting consolidation in the specification, select commodity code to create part number. And introduce how the interface of VPRM and PDMS, how import the weight data, and how make owner part number press in the drawing conveniently. The part numbers are applied in the drawings of LingAo Phrase 2 Nuclear Power Plant, owner accepts them. (authors)

  18. Modified NASA-Lewis Chemical Equilibrium Code for MHD applications

    Energy Technology Data Exchange (ETDEWEB)

    Sacks, R. A.; Geyer, H. K.; Grammel, S. J.; Doss, E. D.

    1979-12-01

    A substantially modified version of the NASA-Lewis Chemical Equilibrium Code has recently been developed. The modifications were designed to extend the power and convenience of the Code as a tool for performing combustor analysis for MHD systems studies. This report describes the effect of the programming details from a user point of view, but does not describe the Code in detail.

  19. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2015-08-01

    Mining is a hazardous industry and high accident rates associated with underground mining is a cause of deep concern. Technological developments notwithstanding, rate of fatal accidents and reportable incidents have not shown corresponding levels of decline. This paper argues that adoption of appropriate safety standards by both mine management and the government may result in appreciable reduction in accident frequency. This can be achieved by using the technology in improving the working conditions, sensitising workers and managers about causes and prevention of accidents. Inputs required for a detailed analysis of an accident include information on location, time, type, cost of accident, victim, nature of injury, personal and environmental factors etc. Such information can be generated from data available in the standard coded accident report form. This paper presents a web based application for accident analysis in Indian mines during 2001-2013. An accident database (SafeStat) prototype based on Intranet of the TCP/IP agreement, as developed by the authors, is also discussed.

  20. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  1. Application of T-Code, Turbo Codes and Pseudo-Random Sequence for Steganography

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2006-01-01

    Full Text Available In this study, we propose new technique that will address the problem of robustness and data safety in steganography. The steganography consists of techniques to allow the communication between two persons, hiding not only the contents but also the very existence of the communication in the eyes of any observer. T-Codes used with Turbo Codes generates cryptic and error-coded data stream, which is hidden in the stego-object using Pseudo-Random sequence. This technique makes our processed data stream non-vulnerable to the attack of an active intruder, or due to noise in the transmission link.

  2. Development of NASA's Accident Precursor Analysis Process Through Application on the Space Shuttle Orbiter

    Science.gov (United States)

    Maggio, Gaspare; Groen, Frank; Hamlin, Teri; Youngblood, Robert

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system. APA docs more than simply track experience: it systematically evaluates experience, looking for under-appreciated risks that may warrant changes to design or operational practice. This paper presents the pilot application of the NASA APA process to Space Shuttle Orbiter systems. In this effort, the working sessions conducted at Johnson Space Center (JSC) piloted the APA process developed by Information Systems Laboratories (ISL) over the last two years under the auspices of NASA's Office of Safety & Mission Assurance, with the assistance of the Safety & Mission Assurance (S&MA) Shuttle & Exploration Analysis Branch. This process is built around facilitated working sessions involving diverse system experts. One important aspect of this particular APA process is its focus on understanding the physical mechanism responsible for an operational anomaly, followed by evaluation of the risk significance of the observed anomaly as well as consideration of generalizations of the underlying mechanism to other contexts. Model completeness will probably always be an issue, but this process tries to leverage operating experience to the extent possible in order to address completeness issues before a catastrophe occurs.

  3. Application of fuzzy decision making to countermeasure strategies after a nuclear accident

    International Nuclear Information System (INIS)

    In the event of a nuclear accident, any decision on countermeasures to protect the public should be made based upon the basic principles recommended by the International Commission on Radiological Protection. The application of these principles requires that there is a balance between the cost and the averted radiation dose, taking into account many subjective factors such as social/political acceptability, psychological stress, and the confidence of the population in the authorities etc. In the framework of classical methods, it is difficult to quantify human subjective judgements and the uncertainties of data efficiently. Hence, any attempt to find the optimal solution for countermeasure strategies without deliberative sensitivity analysis can be misleading. However, fuzzy sets, with linguistic terms to describe the human subjective judgement and with fuzzy numbers to model the uncertainties of the parameters, can be introduced to eliminate these difficulties. With fuzzy rating, a fuzzy multiple attribute decision making method can rank the possible countermeasure strategies. This paper will describe the procedure of the method and present an illustrative example

  4. Current and anticipated uses of thermal-hydraulic codes in NFI

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, K. [Nuclear Fuel Industries, Ltd., Tokyo (Japan); Takayasu, M. [Nuclear Fuel Industries, Ltd., Sennann-gun (Japan)

    1997-07-01

    This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.

  5. Fixed-Length Error Resilient Code and Its Application in Video Coding

    Institute of Scientific and Technical Information of China (English)

    FANChen; YANGMing; CUIHuijuan; TANGKun

    2003-01-01

    Since popular entropy coding techniques such as Variable-length code (VLC) tend to cause severe error propagation in noisy environments, an error resilient entropy coding technique named Fixed-length error resilient code (FLERC) is proposed to mitigate the problem. It is found that even for a non-stationary source, the probability of error propagation could be minimized through introducing intervals into the codeword space of the fixed-length codes. FLERC is particularly suitable for the entropy coding for video signals in error-prone environments, where a little distortion is tolerable, but severe error propagation would lead to fatal consequences. An iterative construction algorithm for FLERC is presented in this paper. In addition, FLERC is adopted instead of VLC as the entropy coder of the DCT coefficients in H.263++Data partitioning slice (DPS) mode, and tested on noisy channels. The simulation results show that this scheme outperforms the scheme of H.263++ combined with FEC when the channel noise is highly extensive, since the error propagation is effectively suppressed by using FLERC. Moreover, it is observed that the reconstructed video quality degrades gracefully as the bit error rate increases.

  6. The development and application of the accident dynamic simulator for dynamic probabilistic risk assessment of nuclear power plants

    International Nuclear Information System (INIS)

    This paper describes the principal modelling concepts, practical aspects, and an application of the Accident Dynamic Simulator (ADS) developed for full scale dynamic probabilistic risk assessment (DPRA) of nuclear power plants. Full scale refers not only to the size of the models, but also to the number of potential sequences which should be studied. Plant thermal-hydraulics behaviour, safety systems response, and operator interactions are explicitly accounted for as integrated active parts in the development of accident scenarios. ADS uses discrete dynamic event trees (D-DET) as the main accident scenario modelling approach, and introduces computational techniques to minimize the computer memory requirement and expedite the simulation. An operator model (including procedure-based behaviour and several types of omission and commission errors) and a thermal-hydraulic model with a PC run time more than 300 times faster than real accident time are among the main modules of ADS. To demonstrate the capabilities of ADS, a dynamic PRA of the Steam Generator Tube Rupture event of a US nuclear power plant is analyzed

  7. 26 CFR 521.102 - Applicable provisions of the Internal Revenue Code.

    Science.gov (United States)

    2010-04-01

    ... 26 Internal Revenue 19 2010-04-01 2010-04-01 false Applicable provisions of the Internal Revenue Code. 521.102 Section 521.102 Internal Revenue INTERNAL REVENUE SERVICE, DEPARTMENT OF THE TREASURY... Revenue Code. (a) The Internal Revenue Code provides in part as follows: Chapter I—Income Tax Sec....

  8. The application of bar coding technology at WIPP

    International Nuclear Information System (INIS)

    Bar coding at the Waste Isolation Pilot Plant (WIPP) can be used to track waste containers within the facility, control transuranic (TRU) waste inventory flow, and reduce manpower and error in recording package identification or control parameters. By choosing where and when to bar code, precise timelines or time-and-motion studies can be conducted to aid in streamlining waste handling throughput at WIPP. Additionally, the use of bar codes as waste container identification (ID) numbers increases the accuracy of recording the ID numbers by four orders of magnitude. The bar code label can also be utilized for other functions, such as shipping labels. Also, the bar code is integrated with the waste management data base such that the entire data base can be accessed on a computer using a bar code

  9. User Effect on Code Application and Qualification Needs

    International Nuclear Information System (INIS)

    Experience with some code assessment case studies and also additional ISPs have shown the dominant effect of the code user on the predicted system behavior. The general findings of the user effect investigations on some of the case studies indicate, specifically, that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under the general title of user effects. The specific characteristics of experimental facilities, i.e. limitations as far as code assessment is concerned; limitations of the used thermal-hydraulic codes to simulate certain system behavior or phenomena; limitations due to interpretation of experimental data by the code user, i.e. interpretation of experimental data base. On the basis of the discussions in this paper, the following conclusions and recommendations can be made: More dialogue appears to be necessary with the experimenters in the planning of code assessment calculations, e.g. ISPs.; User guidelines are not complete for the codes and the lack of sufficient and detailed user guidelines are observed with some of the case studies; More extensive user instruction and training, improved user guidelines, or quality assurance procedures may partially reduce some of the subjective user influence on the calculated results; The discrepancies between experimental data and code predictions are due both to the intrinsic code limit and to the so called user effects. There is a worthful need to quantify the percentage of disagreement due to the poor utilization of the code and due to the code itself. This need especially arises for the uncertainty evaluation studies (e.g. [18]) which do not take into account the mentioned user effects; A much focused investigation, based on the results of comparison calculations e.g. ISPs, analyzing the experimental data and the results of the specific code in order to evaluate the user effects and the related experimental aspects should be integral part of the

  10. QR CODE GENERATION AND APPLICATION OF ANTI-FALSIFICATION

    OpenAIRE

    Ke LIAO

    2016-01-01

    QR code is 2-dimensional bar code. It is more advanced than bar code, which can only store numbers and characters. QR code can store numbers, characters (including Chinese characters) and even images. It has great data capacity and allows for error correction. It is commonly used to spread information and is sometimes used to achieve website login function. It can even be encrypted that the user has to enter password to get the original information. In my thesis, I come up with an idea tha...

  11. Hybrid simulation codes with application to shocks and upstream waves

    Science.gov (United States)

    Winske, D.

    1985-01-01

    Hybrid codes in which part of the plasma is represented as particles and the rest as a fluid are discussed. In the past few years such codes with particle ions and massless, fluid electrons have been applied to space plasmas, especially to collisionless shocks. All of these simulation codes are one-dimensional and similar in structure, except for how the field equations are solved. The various approaches that are used (resistive Ohm's law, predictor-corrector, Hamiltonian) are described in detail and results from the various codes are compared with examples taken from collisionless shocks and low frequency wave phenomena upstream of shocks.

  12. Trends in EFL Technology and Educational Coding: A Case Study of an Evaluation Application Developed on LiveCode

    Science.gov (United States)

    Uehara, Suwako; Noriega, Edgar Josafat Martinez

    2016-01-01

    The availability of user-friendly coding software is increasing, yet teachers might hesitate to use this technology to develop for educational needs. This paper discusses studies related to technology for educational uses and introduces an evaluation application being developed. Through questionnaires by student users and open-ended discussion by…

  13. Modified NASA-Lewis chemical equilibrium code for MHD applications

    Science.gov (United States)

    Sacks, R. A.; Geyer, H. K.; Grammel, S. J.; Doss, E. D.

    1979-01-01

    A substantially modified version of the NASA-Lewis Chemical Equilibrium Code was recently developed. The modifications were designed to extend the power and convenience of the Code as a tool for performing combustor analysis for MHD systems studies. The effect of the programming details is described from a user point of view.

  14. Accident Locations, Currently this layer is maintained by our County Sheriff Department & done with a desktop GIS program., Published in 2013, Not Applicable scale, Chippewa County.

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — This Accident Locations dataset, published at Not Applicable scale, was produced all or in part from Other information as of 2013. It is described as 'Currently...

  15. Application of laser bar code technology in power fitting evaluation

    Science.gov (United States)

    Yang, Xiaohong; Liu, Shuhuab

    2007-12-01

    In this work, an automatic encoding and management system on power fittings (PFEMS) is developed based on laser bar coding technology. The system can encode power fittings according to their types, structure, dimensions, materials, and technical characteristics. Both the character codes and the laser bar codes of power fittings can be produced from the system. The system can evaluate power fittings and search process-paper automatically. The system analyzes the historical values and technical information of congeneric fittings, and forms formulae of evaluation with recursive analytical method. And then stores the formulae and technical documents into the database for index. Scanning the bar code with a laser bar code reader, accurate evaluation and corresponding process-paper of the fittings can be produced. The software has already been applied in some power stations and worked very well.

  16. OCA-P, a deterministic and probabilistic fracture-mechanics code for application to pressure vessels

    International Nuclear Information System (INIS)

    The OCA-P code is a probabilistic fracture-mechanics code that was prepared specifically for evaluating the integrity of pressurized-water reactor vessels when subjected to overcooling-accident loading conditions. The code has two-dimensional- and some three-dimensional-flaw capability; it is based on linear-elastic fracture mechanics; and it can treat cladding as a discrete region. Both deterministic and probabilistic analyses can be performed. For the former analysis, it is possible to conduct a search for critical values of the fluence and the nil-ductility reference temperature corresponding to incipient initiation of the initial flaw. The probabilistic portion of OCA-P is based on Monte Carlo techniques, and simulated parameters include fluence, flaw depth, fracture toughness, nil-ductility reference temperature, and concentrations of copper, nickel, and phosphorous. Plotting capabilities include the construction of critical-crack-depth diagrams (deterministic analysis) and various histograms (probabilistic analysis)

  17. Theory and application of the coded aperture fuel motion detection system

    International Nuclear Information System (INIS)

    A fuel motion detection system based on coded aperture imaging has been developed for the Annular Core Research Reactor. Its configuration evolved after investigations were carried out to determine the required system capabilities. The reactor environment, developments in the theory of coded apertures for nuclear radiations and compatibility with prototypical geometries were considered. The system was fabricated and inserted into the ACRR where it has recorded the fuel motion from a single pin subjected to loss of flow accident conditions. In addition, computer simulations have shown that in the reactor environment and for fast data acquisition, coded imaging, particularly with uniformly redundant arrays, offers significant advantages over pinhole camera geometries. The extension of this technique toward imaging of 37-pin bundles and imaging with fast neutrons is also being investigated

  18. Determination of error probability of cryptography and safety codes for safety-related railway applications

    OpenAIRE

    Maria Franekova; Marek Vyrostko; Peter Luley

    2013-01-01

    The paper deals with the problem of determination of error probability of cryptography and safety codes used within the safety-related railway applications with increasing safety integrity level (SIL). In the paper are also described requirements for cryptographic block code and safety linear block code in safety-related communications for railway application. The main part is oriented to the description of mathematical apparatus for the error probability of the cryptography and safety block ...

  19. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  20. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  1. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  2. Applicability of vector processing to large-scale nuclear codes

    International Nuclear Information System (INIS)

    To meet the growing trend of computational requirements in JAERI, introduction of a high-speed computer with vector processing faculty (a vector processor) is desirable in the near future. To make effective use of a vector processor, appropriate optimization of nuclear codes to pipelined-vector architecture is vital, which will pose new problems concerning code development and maintenance. In this report, vector processing efficiency is assessed with respect to large-scale nuclear codes by examining the following items: 1) The present feature of computational load in JAERI is analyzed by compiling the computer utilization statistics. 2) Vector processing efficiency is estimated for the ten heavily-used nuclear codes by analyzing their dynamic behaviors run on a scalar machine. 3) Vector processing efficiency is measured for the other five nuclear codes by using the current vector processors, FACOM 230-75 APU and CRAY-1. 4) Effectiveness of applying a high-speed vector processor to nuclear codes is evaluated by taking account of the characteristics in JAERI jobs. Problems of vector processors are also discussed from the view points of code performance and ease of use. (author)

  3. Iterative codes and their application in systems for event registration in multichannel charged particle detectors

    International Nuclear Information System (INIS)

    Questions on the use of iterative codes for data registration in hodoscopic systems are considered. The method of construction and properties of the most interesting codes from the practical point of view are considered. These codes can be used to construct effective coding circuits applicable to the systems which have a large number of registration channels (more than one hundred). Examples of the construction of coding circuits for a large number of inputs are given. A long-term application of the iterative codes for the creation of trigger systems used for spectrometers-calorimeters is considered. Efficiency on the use of the iterative codes depending on the number of registration channels is discussed

  4. Transport accident frequency data, their sources and their application in risk assessment

    International Nuclear Information System (INIS)

    Base transport accident frequency data and sources of these data are presented. Both generic information and rates specific to particular routes or packages are included. Strong packages, such as those containing significant quantities of radioactive materials, will survive most of the accidents represented by these base frequencies without a containment breach. The association of severity probability distributions with a base frequency, and package and contents response, leading to the quantification of release frequency and magnitude, are often more important in risk assessment than the base frequency itself. This paper therefore also includes brief comments on techniques adopted to utilize the base frequencies. This paper reports an accident frequency data survey undertaken at the end of 1986. It has not been updated to take account of work published between January 1987 and the Report publication date. (author)

  5. Coded Aperture Imaging for Fluorescent X-rays-Biomedical Applications

    Energy Technology Data Exchange (ETDEWEB)

    Haboub, Abdel; MacDowell, Alastair; Marchesini, Stefano; Parkinson, Dilworth

    2013-06-01

    Employing a coded aperture pattern in front of a charge couple device pixilated detector (CCD) allows for imaging of fluorescent x-rays (6-25KeV) being emitted from samples irradiated with x-rays. Coded apertures encode the angular direction of x-rays and allow for a large Numerical Aperture x- ray imaging system. The algorithm to develop the self-supported coded aperture pattern of the Non Two Holes Touching (NTHT) pattern was developed. The algorithms to reconstruct the x-ray image from the encoded pattern recorded were developed by means of modeling and confirmed by experiments. Samples were irradiated by monochromatic synchrotron x-ray radiation, and fluorescent x-rays from several different test metal samples were imaged through the newly developed coded aperture imaging system. By choice of the exciting energy the different metals were speciated.

  6. Pseudo Quasi-3 Designs and their Applications to Coding Theory

    OpenAIRE

    Bracken, Carl

    2008-01-01

    We define a pseudo quasi-3 design as a symmetric design with the property that the derived and residual designs with respect to at least one block are quasi-symmetric. Quasi-symmetric designs can be used to construct optimal self complementary codes. In this article we give a construction of an infinite family of pseudo quasi-3 designs whose residual designs allow us to construct a family of codes with a new parameter set that meet the Grey Rankin bound.

  7. TIRE MODELS USED IN VEHICLE DYNAMIC APPLICATIONS AND THEIR USING IN VEHICLE ACCIDENT SIMULATIONS

    Directory of Open Access Journals (Sweden)

    Osman ELDOĞAN

    1995-01-01

    Full Text Available Wheel model is very important in vehicle modelling, it is because the contact between vehicle and road is achieved by wheel. Vehicle models can be dynamic models which are used in vehicle design, they can also be models used in accident simulations. Because of the importance of subject, many studies including theoretical, experimental and mixed type have been carried out. In this study, information is given about development of wheel modelling and research studies and also use of these modellings in traffic accident simulations.

  8. Advance of Hazardous Operation Robot and its Application in Special Equipment Accident Rescue

    Science.gov (United States)

    Zeng, Qin-Da; Zhou, Wei; Zheng, Geng-Feng

    A survey of hazardous operation robot is given out in this article. Firstly, the latest researches such as nuclear industry robot, fire-fighting robot and explosive-handling robot are shown. Secondly, existing key technologies and their shortcomings are summarized, including moving mechanism, control system, perceptive technology and power technology. Thirdly, the trend of hazardous operation robot is predicted according to current situation. Finally, characteristics and hazards of special equipment accident, as well as feasibility of hazardous operation robot in the area of special equipment accident rescue are analyzed.

  9. The Genomic Code: Genome Evolution and Potential Applications

    KAUST Repository

    Bernardi, Giorgio

    2016-01-25

    The genome of metazoans is organized according to a genomic code which comprises three laws: 1) Compositional correlations hold between contiguous coding and non-coding sequences, as well as among the three codon positions of protein-coding genes; these correlations are the consequence of the fact that the genomes under consideration consist of fairly homogeneous, long (≥200Kb) sequences, the isochores; 2) Although isochores are defined on the basis of purely compositional properties, GC levels of isochores are correlated with all tested structural and functional properties of the genome; 3) GC levels of isochores are correlated with chromosome architecture from interphase to metaphase; in the case of interphase the correlation concerns isochores and the three-dimensional “topological associated domains” (TADs); in the case of mitotic chromosomes, the correlation concerns isochores and chromosomal bands. Finally, the genomic code is the fourth and last pillar of molecular biology, the first three pillars being 1) the double helix structure of DNA; 2) the regulation of gene expression in prokaryotes; and 3) the genetic code.

  10. QR Code 二维码在准考证中的应用%Application of QR code two - dimensional code in the examination certificate

    Institute of Scientific and Technical Information of China (English)

    李文江; 陈诗琴

    2015-01-01

    针对现有纸质准考证身份验证存在的缺点,基于 QR Code 二维码技术,设计并实现了准考证的二维码生成、发布与识别,并从需求分析与技术思路、应用设计、具体实现和实际运行效果等方面进行阐述。二维码的应用可以极大地方便考试组织部门和考生。%Based on the QR Code technology,this paper designs and implements the confirmation code generation,distribution and identification according to the disadvantages of the existing paper ticket authenti-cation,and the paper explains in detail from the aspects of demand analysis and technical ideas,application design,implementation and practical operation effect. The application of two - dimensional code promotes the fairness of the exam. Also it greatly facilitates the examination of the organization department and the candidates.

  11. Experimental study of the interplay of channel and network coding in low power sensor applications

    OpenAIRE

    Angelopoulos, Georgios; Paidimarri, Arun; Chandrakasan, Anantha P.; Medard, Muriel

    2013-01-01

    In this paper, we evaluate the performance of random linear network coding (RLNC) in low data rate indoor sensor applications operating in the ISM frequency band. We also investigate the results of its synergy with forward error correction (FEC) codes at the PHY-layer in a joint channel-network coding (JCNC) scheme. RLNC is an emerging coding technique which can be used as a packet-level erasure code, usually implemented at the network layer, which increases data reliability against channel f...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

  13. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  14. Development of methods for the analysis of accident scenarios with steam line breaks and boron dilution by the help of the code system ATHLET-DYN3D. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Libraries of two-group neutron-diffusion parameters for a Siemens-KWU-Konvoi Pressurized Water Reactor have been generated at Forschungszentrum Rossendorf and TUeV Bau und Betrieb GmbH by using the codes HELIOS and CASMO, respectively. The libraries have been coupled to the reactor-dynamics code DYN3D. For a generic PWR core containing MOX fuel elements, DYN3D macro-burnup calculations and the calculation of different operation states have been carried out. The results will be used for the investigation of possible accident scenarios. Reactivity coefficients calculated by DYN3D are needed for accident analyses by the 1-D thermal-hydraulic code ATHLET. Using the cross section data, more detailed analyses can be carried out by applying the coupled-code system DYN3D-ATHLET, considering 3D neutron kinetics. The comparison of the results calculated by DYN3D with two different diffusion-parameter libraries can give an idea of how uncertainties in diffusion data influence the accuracy of reactor simulation. (orig.)

  15. Incorporation of phenomenological uncertainties in probabilistic safety analysis - application to LMFBR core disruptive accident energetics

    Energy Technology Data Exchange (ETDEWEB)

    Najafi, B; Theofanous, T G; Rumble, E T; Atefi, B

    1984-08-01

    This report describes a method for quantifying frequency and consequence uncertainty distribution associated with core disruptive accidents (CDAs). The method was developed to estimate the frequency and magnitude of energy impacting the reactor vessel head of the Clinch River Breeder Plant (CRBRP) given the occurrence of hypothetical CDAs. The methodology is illustrated using the CRBR example.

  16. Application of GIS in prediction and assessment system of off-site accident consequence for NPP

    International Nuclear Information System (INIS)

    The assessment and prediction software system of off-site accident consequence for Guangdong Nuclear Power Plant (GNARD2.0) is a GIS-based software system. The spatial analysis of radioactive materials and doses with geographic information is available in this system. The structure and functions of the GNARD system and the method of applying ArcView GIS are presented

  17. Wavelet image coding with parametric thresholding: application to JPEG2000

    Science.gov (United States)

    Zaid, Azza O.; Olivier, Christian; Marmoiton, Francois

    2003-05-01

    With the increasing use of multimedia technologies, image compression requires higher performance as well as new features. To address this need in the specific area of image coding, the latest ISO/IEC image compression standard, JPEG2000, has been developed. In part II of the standard, the Wavelet Trellis Coded Quantization (WTCQ) algorithm was adopted. It has been proved that this quantization design provides subjective image quality superior to other existing quantization techniques. In this paper we are aiming to improve the rate-distortion performance of WTCQ, by incorporating a thresholding process in JPEG2000 coding chain. The threshold decisions are derived in a Bayesian framework, and the prior used on the wavelet coefficients is the generalized Gaussian distribution (GGD). The threshold value depends on the parametric model estimation of the subband wavelet coefficient distribution. Our algorithm approaches the lowest possible memory usage by using line-based wavelet transform and a scan-based bit allocation technique. In our work, we investigate an efficient way to apply the TCQ to wavelet image coding with regard to both the computational complexity and the compression performance. Experimental results show that the proposed algorithm performs competitively with the best available coding algorithms reported in the literature in terms quality performance.

  18. Rising and boiling of a drop of volatile liquid in a heavier one: application to the LMFBR severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Pigny, Sylvain L.; Coste, Pierre F. [DEN/DER/SSTH, CEA/Grenoble, 38054 Grenoble Cedex 9 (France)

    2005-07-01

    Full text of publication follows: The rising and, simultaneously the boiling, of a droplet of volatile liquid in a heavier one is computation-ally investigated. Our calculations are performed with the help of the SIMMER code, in which a specific DNS algorithm is developed, to represent surface tension between the different media in an explicit way. This is required to represent the physical contact that occurs between two liquids and the vapor from the lighter one, since interfacial heat transfers, and therefore boiling kinetics, merely depend on it. The behavior of the three fluids system is of interest as a key phenomenon related to the transition phase of LMFBR severe accidents, before the formation of a fully developed bubble column. The driven force due to the boiling of steel drops can play a major role in the relocation, and, consequently, the recriticality of UO{sub 2} fuel. The problem is investigated focusing first on analytical experiments, built-up with simulating materials, and for which accurate experimental results are provided. The dependence of results with regard to thermodynamical and physical properties is underlined. This point is of interest in view of some uncertainties in the knowledge of data concerning the materials present in the reactor at high temperature. The pressure level is a key parameter in the accident scenarios: its influence is uppermost on the volumic mass of the gas. It is also outlined. (authors)

  19. MELCOR computer code manuals

    International Nuclear Information System (INIS)

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

  20. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

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

  1. Potential application of item-response theory to interpretation of medical codes in electronic patient records

    Directory of Open Access Journals (Sweden)

    Dregan Alex

    2011-12-01

    Full Text Available Abstract Background Electronic patient records are generally coded using extensive sets of codes but the significance of the utilisation of individual codes may be unclear. Item response theory (IRT models are used to characterise the psychometric properties of items included in tests and questionnaires. This study asked whether the properties of medical codes in electronic patient records may be characterised through the application of item response theory models. Methods Data were provided by a cohort of 47,845 participants from 414 family practices in the UK General Practice Research Database (GPRD with a first stroke between 1997 and 2006. Each eligible stroke code, out of a set of 202 OXMIS and Read codes, was coded as either recorded or not recorded for each participant. A two parameter IRT model was fitted using marginal maximum likelihood estimation. Estimated parameters from the model were considered to characterise each code with respect to the latent trait of stroke diagnosis. The location parameter is referred to as a calibration parameter, while the slope parameter is referred to as a discrimination parameter. Results There were 79,874 stroke code occurrences available for analysis. Utilisation of codes varied between family practices with intraclass correlation coefficients of up to 0.25 for the most frequently used codes. IRT analyses were restricted to 110 Read codes. Calibration and discrimination parameters were estimated for 77 (70% codes that were endorsed for 1,942 stroke patients. Parameters were not estimated for the remaining more frequently used codes. Discrimination parameter values ranged from 0.67 to 2.78, while calibration parameters values ranged from 4.47 to 11.58. The two parameter model gave a better fit to the data than either the one- or three-parameter models. However, high chi-square values for about a fifth of the stroke codes were suggestive of poor item fit. Conclusion The application of item response

  2. An Improved Real-Coded Genetic Algorithm and Its Application

    Institute of Scientific and Technical Information of China (English)

    Zhong-Lai Wang; Ping Yang; Dan Ling; Qiang Miao

    2008-01-01

    Real-coded genetic algorithm (RGA) usuaUy meets the demand of consecutive space problem. However, compared with simple genetic algorithm (SGA), RGA also has the inherent disadvantages such as prematurity and slow convergence when the solution is close to the optimum solution. This paper presents an improved real-coded genetic algorithm to increase the computation efficiency and avoid prematurity, especially in the optimization of multi-modal function. In this method, mutation operation and crossover operation are improved. Examples are given to demonstrate its computation efficiency and robustness.

  3. [The QR code in society, economy and medicine--fields of application, options and chances].

    Science.gov (United States)

    Flaig, Benno; Parzeller, Markus

    2011-01-01

    2D codes like the QR Code ("Quick Response") are becoming more and more common in society and medicine. The application spectrum and benefits in medicine and other fields are described. 2D codes can be created free of charge on any computer with internet access without any previous knowledge. The codes can be easily used in publications, presentations, on business cards and posters. Editors choose between contact details, text or a hyperlink as information behind the code. At expert conferences, linkage by QR Code allows the audience to download presentations and posters quickly. The documents obtained can then be saved, printed, processed etc. Fast access to stored data in the internet makes it possible to integrate additional and explanatory multilingual videos into medical posters. In this context, a combination of different technologies (printed handout, QR Code and screen) may be reasonable. PMID:21805904

  4. PROSA-1: a probabilistic response-surface analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vaurio, J. K.; Mueller, C.

    1978-06-01

    Techniques for probabilistic response-surface analysis have been developed to obtain the probability distributions of the consequences of postulated nuclear-reactor accidents. The uncertainties of the consequences are caused by the variability of the system and model input parameters used in the accident analysis. Probability distributions are assigned to the input parameters, and parameter values are systematically chosen from these distributions. These input parameters are then used in deterministic consequence analyses performed by mechanistic accident-analysis codes. The results of these deterministic consequence analyses are used to generate the coefficients for analytical functions that approximate the consequences in terms of the selected input parameters. These approximating functions are used to generate the probability distributions of the consequences with random sampling being used to obtain values for the accident parameters from their distributions. A computer code PROSA has been developed for implementing the probabilistic response-surface technique. Special features of the code generate or treat sensitivities, statistical moments of the input and output variables, regionwise response surfaces, correlated input parameters, and conditional distributions. The code can also be used for calculating important distributions of the input parameters. The use of the code is illustrated in conjunction with the fast-running accident-analysis code SACO to provide probability studies of LMFBR hypothetical core-disruptive accidents. However, the methods and the programming are general and not limited to such applications.

  5. Application of computer code ALMOD in transient analysis with reverse flow in the primary loop of nuclear power plant

    International Nuclear Information System (INIS)

    A computer code ALMOD 3W3 to analyze the transients in which reverse flow in the primary loop of nuclear power plant may occur has been developed. The method to calculate the fluid dynamics in NRC system is presented. The locked rotor accident in one coolant loop is analyzed. (author)

  6. Experimental Study of Application Specific Source Coding for Wireless Sensor Networks

    OpenAIRE

    Annamalai, Muthiah; Shrestha, Darshan; Tjuatja, Saibun

    2008-01-01

    The energy bottleneck in Wireless Sensor Network(WSN) can be reduced by limiting communication overhead. Application specific source coding schemes for the sensor networks provide fewer bits to represent the same amount of information exploiting the redundancy present in the source model, network architecture and the physical process. This paper reports the performance of representative codes from various families of source coding schemes (lossless, lossy, constant bit-rate, variable bit-rate...

  7. Improved predictions of nuclear reaction rates with the TALYS reaction code for astrophysical applications

    OpenAIRE

    Goriely, S.; Hilaire, S; Koning, A.J.

    2008-01-01

    Nuclear reaction rates of astrophysical applications are traditionally determined on the basis of Hauser-Feshbach reaction codes. These codes adopt a number of approximations that have never been tested, such as a simplified width fluctuation correction, the neglect of delayed or multiple-particle emission during the electromagnetic decay cascade, or the absence of the pre-equilibrium contribution at increasing incident energies. The reaction code TALYS has been recently updated to estimate t...

  8. Partial-MDS Codes and their Application to RAID Type of Architectures

    CERN Document Server

    Blaum, Mario; Hetzler, Steven

    2012-01-01

    A family of codes with a natural two-dimensional structure is presented, inspired by an application of RAID type of architectures whose units are solid state drives (SSDs). Arrays of SSDs behave differently to arrays of hard disk drives (HDDs), since hard errors in sectors are common and traditional RAID approaches (like RAID 5 or RAID 6) may be either insufficient or excessive. An efficient solution to this problem is given by the new codes presented, called partial-MDS (PMDS) codes.

  9. Validation of the MORET 5 code for criticality safety applications

    International Nuclear Information System (INIS)

    The MORET-5 Monte Carlo code includes 2 calculation routes: a multi-group route based on cross-sections calculated from various cell codes such a APOLLO2, DRAGON4 or SCALE, and a continuous energy calculation route. The validation of the MORET-5 code is done through the comparison between the calculated benchmark k(eff) and the experimental benchmark k(eff). If the discrepancy between these 2 k(eff) is higher than the combined standard deviation of the benchmark uncertainty and the Monte Carlo standard deviation, a bias can be identified. The criticality experimental validation database is made up of 2255 benchmarks. Concerning the multi-group approach, the present work deals only with the APOLLO2 - MORET-5 route. The APOLLO2 cell code uses a 281 energy-group structure library based on JEFF3.1. Preliminary analyses have shown that the continuous energy route using JEFF3.1 or ENDF/B-VII.0 libraries are in good agreement with the experimental k(eff) in the majority of cases. Regarding the APOLLO2 - MORET-5 calculation route, some improvements are still needed, especially for what concerns the multi-group treatment

  10. CONTAIN-LMR程序中池式钠火事故分析计算模型的验证%Verification of sodium pool fire accident analysis model in CONTAIN-LMR code

    Institute of Scientific and Technical Information of China (English)

    李世锐; 任丽霞; 胡文军; 乔鹏瑞

    2016-01-01

    CONTAIN-LMR是针对以液态钠为冷却剂的反应堆而开发的安全壳事故一体化分析程序。我国目前的CONTAIN-LMR程序版本为2000年左右从法国引进,还未进行过面向工程设计的系统性地程序开发和验证。本文主要针对 CONTAIN-LMR 程序中模拟池式钠火事故的分析模型进行详细分析,并采用国际上的池式钠火实验进行验证,实验验证结果表明 CONTAIN-LMR 程序可以较准确地模拟池式钠火事故造成的钠工艺间内的温度、压力升高及放射性钠气溶胶行为。本文的研究结果初步表明CONTAIN-LMR程序可用于钠冷快堆的钠火事故分析。%CONTAIN-LMR is an integrated code which aims at sodium cooled fast reactor containment accident analysis. The current version of the CONTAIN-LMR code in China was imported from France around 2000,program development and verification of engineering level design has not undertaken systematically. This paper makes a detailed analysis for the models of sodium pool fire accident simulation in CONTAIN-LMR code,and uses international sodium pool fire experiments for verification,the result shows that the CONTAIN-LMR code can simulate the temperature,pressure rising and radioactive sodium aerosol behavior in containment caused by sodium pool fire accidents. The studies in this paper indicated that the CONTAIN-LMR code can be used for the analysis of sodium fire accidents in sodium cooled fast reactor.

  11. Python-Assisted MODFLOW Application and Code Development

    Science.gov (United States)

    Langevin, C.

    2013-12-01

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

  12. Development and application of a radioactivity evaluation technique the to obtain radiation exposure dose of radioactivity evaluation technique when a severe accident occurs in the a power station of a severe accident. Accident management guidelines of knowledge-based maintenance

    International Nuclear Information System (INIS)

    As a One of the lessons learned from the nuclear accident at the Fukushima Daiichi Nuclear Power Stations of Tokyo Electric Power Company, the was the need for improvement of accident management guidelines is required. In this report study, we developed and applied a dose evaluation technique to evaluated the radiation dose in a nuclear power plant assuming three conditions: employees were evacuation evacuated at the time of a severe accident occurrence; operators carried out the accident management operation; of the operators, and the repair work was carried out for of the trouble damaged apparatuses in a the nuclear power plant using a dose evaluation system. The following knowledge findings were obtained and should to be reflected to in the knowledge base of the guidelines was obtained. (1) By making clearly identifying an areas beforehand becoming the that would receive high radiation doses at the time of a severe accident definitely beforehand, we can employees can be moved to the evacuation places through an areas having of low dose rate and it is also known it how much we long employees can safely stay in the evacuation places. (2) When they circulate CV containment vessel recirculation sump water is recirculated by for the accident management operation and the restoration of safety in the facilities, because the plumbing piping and the apparatuses become radioactive radioactivity sources, the dose evaluation of the shortest access route and detour access routes with should be made for effective the accident management operation is effective. Because the area where a dose rate rises changes which as safety apparatuses are restored, in consideration of a plant state, it is necessary to judge the rightness or wrongness of the work continuation from the spot radioactive dose of the actual apparatus area, with based on precedence of the need to restore with precedence, and to choose a system to be used for accident management. (author)

  13. Enhancing the scope of applications of standard hydraulic codes by linking with others

    International Nuclear Information System (INIS)

    In the nuclear plant safety analysis various computer codes are available which simulate the overall plant response or the behavior of the single systems or components. Each of these codes is tailor-made, so as to match the requirements of it's scope of applications. The general purpose code, combining the advantages of existing individual ones, will not be available in the near future since the development costs of such a code will widely exceed the benefits to be yielded. Hence, more reasonable and more economical seems to be the linking of the already existing, verified and approved codes individually, following the modular-design principle. The paper deals with the philosophy of developing the Code-Systems, using the software interfaces allowing to link external individual Component-Codes to the standard thermal hydraulic codes, like RELAP5, TRAC or CATHENA, in order to enhance the scope of applications far beyond their present capabilities. The need and the purpose of such virtual coupling are approved and illustrated from the general point of view. The main groups of applications where the procedure of linking with external specialized codes gained more benefits than off-line iterations are described. In addition, the state-of-the-art with appropriate comprehensive overview of the literature dealing with already approved coupling is presented

  14. Development of an Ontology to Assist the Modeling of Accident Scenarii "Application on Railroad Transport "

    CERN Document Server

    Maalel, Ahmed; Mejri, Lassad; Ghezela, Henda Hajjami Ben

    2012-01-01

    In a world where communication and information sharing are at the heart of our business, the terminology needs are most pressing. It has become imperative to identify the terms used and defined in a consensual and coherent way while preserving linguistic diversity. To streamline and strengthen the process of acquisition, representation and exploitation of scenarii of train accidents, it is necessary to harmonize and standardize the terminology used by players in the security field. The research aims to significantly improve analytical activities and operations of the various safety studies, by tracking the error in system, hardware, software and human. This paper presents the contribution of ontology to modeling scenarii for rail accidents through a knowledge model based on a generic ontology and domain ontology. After a detailed presentation of the state of the art material, this article presents the first results of the developed model.

  15. Improvement on reaction model for sodium-water reaction jet code and application analysis

    International Nuclear Information System (INIS)

    In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3 (SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated. (author)

  16. Improvement on reaction model for sodium-water reaction jet code and application analysis

    Energy Technology Data Exchange (ETDEWEB)

    Itooka, Satoshi; Saito, Yoshinori [Hitachi Ltd., Nuclear Systems Division, Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Murata, Shuuichi [Hitachi Engineering Co., Ltd., Nuclear Power Plant Engineering No.2 Dept., Hitachi, Ibaraki (Japan)

    2000-03-01

    In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3{center_dot}Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3 (SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated. (author)

  17. Event trees and dynamic event trees: Applications to steam generator tube rupture accidents

    International Nuclear Information System (INIS)

    The dynamic event tree analysis method (DETAM) is a simulation based approach that models the integrated, dynamic response of the plant/operating crew system to an accident. It extends the conventional event tree/fault tree methodology for accident sequence analysis in two ways. First, it allows for tree branchings at discrete points in time. Second, the tree sequences explicitly track changes in the operating crew state, as well as changes in the plant hardware state. Process variable calculations and operating procedures are used in linking the crew and hardware behaviour. - The paper compares the conventional event tree/fault tree methodology for accident sequence analysis with the dynamic event tree method in the analysis of a pressurized water reactor steam generator tube rupture. Two previous PSA analyses are used for the comparison. The first employs the ''event tree with boundary conditions'' approach and uses fairly detailed top event headings. The second employs the ''linked fault tree'' approach and uses a relatively small event tree. - A quantitative comparison of the results of the three analyses shows that, in this particularly case study, the DETAM results appear to be less conservative. This is due, in part, to DETAM's treatment of recovery actions embedded in the emergency operating procedures. The quantitative results, however, should be viewed with some caution, since: (a) the three analyses have different scopes and employ different assumptions, and (b) a number of the parameters used in the DETAM analysis are highly uncertain. - A qualitative comparison of results shows that the dominant sequences predicted by each methodology are similar. However, the DETAM scenario descriptions are more detailed and allow better definition of steps to reduce risk. Further, the DETAM models deal with the variety of human error forms and their consequences; this provides a better capability of identifying and quantifying complex accident scenarios that may not

  18. Recent Developments in Level 2 PSA and Severe Accident Management

    International Nuclear Information System (INIS)

    In 1997, CSNI WGRISK produced a report on the state of the art in Level 2 PSA and severe accident management - NEA/CSNI/R(1997)11. Since then, there have been significant developments in that more Level 2 PSAs have been carried out worldwide for a variety of nuclear power plant designs including some that were not addressed in the original report. In addition, there is now a better understanding of the severe accident phenomena that can occur following core damage and the way that they should be modelled in the PSA. As requested by CSNI in December 2005, the objective of this study was to produce a report that updates the original report and gives an account of the developments that have taken place since 1997. The aim has been to capture the most significant new developments that have occurred rather than to provide a full update of the original report, most of which is still valid. This report is organised using the same structure as the original report as follows: Chapter 2: Summary on state of application, results and insights from recent Level 2 PSAs. Chapter 3: Discussion on key severe accident phenomena and modelling issues, identification of severe accident issues that should be treated in Level 2 PSAs for accident management applications, review of severe accident computer codes and the use of these codes in Level 2 PSAs. Chapter 4: Review of approaches and practices for accident management and SAM, evaluation of actions in Level 2 PSAs. Chapter 5: Review of available Level 2 PSA methodologies, including accident progression event tree / containment event tree development. Chapter 6: Aspects important to quantification, including the use of expert judgement and treatment of uncertainties. Chapter 7: Examples of the use of the results and insights from the Level 2 PSA in the context of an integrated (risk informed) decision making process

  19. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    International Nuclear Information System (INIS)

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  20. Application of CFD Code PHOENICS for simulating CYCLONE SEPARATORS

    International Nuclear Information System (INIS)

    The work presents a computational fluid dynamics (CFD) calculation to investigate the flow field in a tangential inlet cyclone which is mainly used for the separation of the moisture from an air stream. Three-dimensional, steady state Eulerian simulations of the turbulent gas - droplet flow in a cyclone separator have been performed. Numerical simulation was carried out using CFD code PHOENICS for the given geometry of separators available in literature

  1. WASA-BOSS. ATHLET-CD model for severe accident analysis for a generic KONVOI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tusheva, Polina; Schaefer, Frank; Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany). Reactor Safety Div.; Hollands, Thorsten [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Trometer, Ailine; Buck, Michael [Stuttgart Univ. (Germany). Dept. of Reactor Safety, Systems and Environment

    2015-07-15

    Within the scope of the ongoing joint research project WASA-BOSS (Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Stoerfallmassnahmen) an ATHLET-CD model for investigation of severe accident scenarios has been developed. The model represents a generic pressurized water reactor (PWR) of type KONVOI. It has been applied for analyzing selected hypothetical core degradation scenarios, considering application of countermeasures and accident management measures, during the early phase of an accident, as well as the late in-vessel phase, when the core degradation process has already begun. Possible accident management measures for loss of coolant (LOCA) and station blackout (SBO) scenarios are discussed. This paper focuses on the ATHLET-CD model development and results from selected simulations for a SBO scenario without and with application of countermeasures.

  2. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    International Nuclear Information System (INIS)

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level

  3. Guidelines and procedures for the International Code Assessment and Applications Program

    International Nuclear Information System (INIS)

    This document presents the guidelines and procedures by which the International Code Assessment and Applications Program (ICAP) will be conducted. The document summarizes the management structure of the program and the relationships between and responsibilities of the United States Nuclear Regulatory Commission (USNRC) and the international participants. The procedures for code maintenance and necessary documentation are described. Guidelines for the performance and documentation of code assessment studies are presented. An overview of an effort to quantify code uncertainty, which the ICAP supports, is included

  4. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  5. Codes of conduct in the Swedish business sector: Application on external parties

    OpenAIRE

    Daniels, Erik; Ryman, Adam

    2013-01-01

    This qualitative study was conducted to investigate how large publicly listed companies inSweden apply its code of conduct on external parties as well as how such application is bestcarried out. Three overarching steps (implementation, monitoring and actions) weredeveloped from previous research to serve this purpose. Agency and contract theory wereimportant to understand the underlying problems of application of codes of conduct onexternal parties. It was found that companies that adequately...

  6. Nuclear accidents

    International Nuclear Information System (INIS)

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  7. Bicycle accidents.

    Science.gov (United States)

    Lind, M G; Wollin, S

    1986-01-01

    Information concerning 520 bicycle accidents and their victims was obtained from medical records and the victims' replies to questionnaires. The analyzed aspects included risk of injury, completeness of accident registrations by police and in hospitals, types of injuries and influence of the cyclists' age and sex, alcohol, fatigue, hunger, haste, physical disability, purpose of cycling, wearing of protective helmet and other clothing, type and quality of road surface, site of accident (road junctions, separate cycle paths, etc.) and turning manoeuvres.

  8. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  9. Experimental Study of Application Specific Source Coding for Wireless Sensor Networks

    CERN Document Server

    Annamalai, Muthiah; Tjuatja, Saibun

    2008-01-01

    The energy bottleneck in Wireless Sensor Network(WSN) can be reduced by limiting communication overhead. Application specific source coding schemes for the sensor networks provide fewer bits to represent the same amount of information exploiting the redundancy present in the source model, network architecture and the physical process. This paper reports the performance of representative codes from various families of source coding schemes (lossless, lossy, constant bit-rate, variable bit-rate, distributed and joint encoding/decoding) in terms of energy consumed, bit-rate achieved, quantization-error/reconstruction-error, latency and complexity of encoder-decoder(codec). A reusable frame work for testing source codes is provided. Finally we propose a set of possible applications and suitable source codes in terms of these parameters.

  10. A new trend in Sabancı University Information Center: QR code application

    OpenAIRE

    Özel, Cem; Ozel, Cem; Akkurt, Mine

    2014-01-01

    The rapid development of mobile technologies in recent years has facilitated the use of the popular QR code application for various purposes. The new generation’s rapid adaptation to change has allowed this application's widespread usage. QR codes with structural properties can be supported with new ideas. It has developed into a new trend in libraries/information centers, as well as in the other areas. One of the usage areas of the QR code is in the marketing field. In this study, various QR...

  11. Tools for signal compression applications to speech and audio coding

    CERN Document Server

    Moreau, Nicolas

    2013-01-01

    This book presents tools and algorithms required to compress/uncompress signals such as speech and music. These algorithms are largely used in mobile phones, DVD players, HDTV sets, etc. In a first rather theoretical part, this book presents the standard tools used in compression systems: scalar and vector quantization, predictive quantization, transform quantization, entropy coding. In particular we show the consistency between these different tools. The second part explains how these tools are used in the latest speech and audio coders. The third part gives Matlab programs simulating t

  12. Application of the RESRAD computer code to VAMP scenario S

    International Nuclear Information System (INIS)

    The RESRAD computer code developed at Argonne National Laboratory was among 11 models from 11 countries participating in the international Scenario S validation of radiological assessment models with Chernobyl fallout data from southern Finland. The validation test was conducted by the Multiple Pathways Assessment Working Group of the Validation of Environmental Model Predictions (VAMP) program coordinated by the International Atomic Energy Agency. RESRAD was enhanced to provide an output of contaminant concentrations in environmental media and in food products to compare with measured data from southern Finland. Probability distributions for inputs that were judged to be most uncertain were obtained from the literature and from information provided in the scenario description prepared by the Finnish Centre for Radiation and Nuclear Safety. The deterministic version of RESRAD was run repeatedly to generate probability distributions for the required predictions. These predictions were used later to verify the probabilistic RESRAD code. The RESRAD predictions of radionuclide concentrations are compared with measured concentrations in selected food products. The radiological doses predicted by RESRAD are also compared with those estimated by the Finnish Centre for Radiation and Nuclear Safety

  13. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  14. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  15. Advancements and performance of iterative methods in industrial applications codes on CRAY parallel/vector supercomputers

    Energy Technology Data Exchange (ETDEWEB)

    Poole, G.; Heroux, M. [Engineering Applications Group, Eagan, MN (United States)

    1994-12-31

    This paper will focus on recent work in two widely used industrial applications codes with iterative methods. The ANSYS program, a general purpose finite element code widely used in structural analysis applications, has now added an iterative solver option. Some results are given from real applications comparing performance with the tradition parallel/vector frontal solver used in ANSYS. Discussion of the applicability of iterative solvers as a general purpose solver will include the topics of robustness, as well as memory requirements and CPU performance. The FIDAP program is a widely used CFD code which uses iterative solvers routinely. A brief description of preconditioners used and some performance enhancements for CRAY parallel/vector systems is given. The solution of large-scale applications in structures and CFD includes examples from industry problems solved on CRAY systems.

  16. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  17. A Synchronization Algorithm and Implementation for High-Speed Block Codes Applications. Part 4

    Science.gov (United States)

    Lin, Shu; Zhang, Yu; Nakamura, Eric B.; Uehara, Gregory T.

    1998-01-01

    Block codes have trellis structures and decoders amenable to high speed CMOS VLSI implementation. For a given CMOS technology, these structures enable operating speeds higher than those achievable using convolutional codes for only modest reductions in coding gain. As a result, block codes have tremendous potential for satellite trunk and other future high-speed communication applications. This paper describes a new approach for implementation of the synchronization function for block codes. The approach utilizes the output of the Viterbi decoder and therefore employs the strength of the decoder. Its operation requires no knowledge of the signal-to-noise ratio of the received signal, has a simple implementation, adds no overhead to the transmitted data, and has been shown to be effective in simulation for received SNR greater than 2 dB.

  18. Tilings with $n$-Dimensional Chairs and their Applications to WOM Codes

    CERN Document Server

    Buzaglo, Sarit

    2012-01-01

    An $n$-dimensional chair consists of an $n$-dimensional box from which a smaller $n$-dimensional box is removed. A tiling of an $n$-dimensional chair has two nice applications in coding for write-once memories. The first one is in the design of codes which correct asymmetric errors with limited-magnitude. The second one is in the design of $n$ cells $q$-ary write-once memory codes. We show an equivalence between the design of a tiling with an integer lattice and the design of a tiling from a generalization of splitting (or of Sidon sequences). A tiling of an $n$-dimensional chair can define a perfect code for correcting asymmetric errors with limited-magnitude. We present constructions for such tilings and prove cases where perfect codes for these type of errors do not exist.

  19. Statistical method application to knowledge base building for reactor accident diagnostic system

    International Nuclear Information System (INIS)

    In the development of a knowledge based expert system, one of key issues is how to build the knowledge base (KB) in an efficient way with keeping the objectivity of KB. In order to solve this issue, an approach has been proposed to build a prototype KB systematically by a statistical method, factor analysis. For the verification of this approach, factor analysis was applied to build a prototype KB for the JAERI expert system DISKET. To this end, alarm and process information was generated by a PWR simulator and the factor analysis was applied to this information to define taxonomy of accident hypotheses and to extract rules for each hypothesis. The prototype KB thus built was tested through inferring against several types of transients including double-failures. In each diagnosis, the transient type was well identified. Furthermore, newly introduced standards for rule extraction showed good effects on the enhancement of the performance of prototype KB. (author)

  20. Radiative Transfer Code: Application to the calculation of PAR

    Indian Academy of Sciences (India)

    D Emmanuel; D Phillippe; C Malik

    2000-12-01

    The production of carbon in the ocean, the so-called primary production, depends on various physico- biological parameters: the biomass and nutrient amounts in oceans, the salinity and temperature of the water and the light available in the water column. We focus on the visible spectrum of the solar radiation defined as the Photosynthetically Active Radiation (PAR). We developed a model (Chami et al. 1997) to simulate the behavior of the solar beam in the atmosphere and the ocean. We first describe the theoretical basis of the code and the method we used to solve the radiative transfer equation (RTE): the successive orders of scattering (SO). The second part deals with a sensitivity study of the PAR just above and below the sea surface for various atmospheric conditions. In a cloudy sky, we computed a ratio between vector fluxes just above the sea surface and spherical fluxes just beneath the sea surface. When the optical thickness of the cloud increases this ratio remains constant and around 1.29. This parameter is convenient to convert vector flux at the sea surface as retrieved from satellite to PAR. Subsequently, we show how solar radiation as vector flux rather than PAR leads to an underestimate of the primary production up to 40% for extreme cases.

  1. Code scaling applicability to a cold leg SBLOCA scenario in a nuclear power plant

    International Nuclear Information System (INIS)

    The knowledge of thermalhydraulic phenomena occurring in Nuclear Power Plants (NPPs) during an accident is crucial in the assessment of nuclear safety. Several experimental facilities have been built to reproduce some accidental scenarios obtaining measured data to be compared with simulation results and to test the capability of the thermalhydraulic codes. This paper presents the scaling method used to obtain a NPP scaled-up TRACE5 input from a Large Scale Test Facility (LSTF) model to reproduce a cold leg Small Break Loss-of-Coolant Accident (SBLOCA) transient. The scaled-up TRACE5 model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. A comparison of the simulation results between LSTF and NPP scaled-up TRACE5 models is provided throughout different graphs, which represent the main thermalhydraulic variables. Results show that the main physical phenomena produced during a cold leg SBLOCA are qualitatively reproduced with the NPP scaled TRACE5 model. (author)

  2. Joint Rate Selection and Wireless Network Coding for Time Critical Applications

    CERN Document Server

    Wang, Xiumin; Xu, Yinlong

    2011-01-01

    In this paper, we dynamically select the transmission rate and design wireless network coding to improve the quality of services such as delay for time critical applications. With low transmission rate, and hence longer transmission range, more packets may be encoded together, which increases the coding opportunity. However, low transmission rate may incur extra transmission delay, which is intolerable for time critical applications. We design a novel joint rate selection and wireless network coding (RSNC) scheme with delay constraint, so as to minimize the total number of packets that miss their deadlines at the destination nodes. We prove that the proposed problem is NPhard, and propose a novel graph model and transmission metric which consider both the heterogenous transmission rates and the packet deadline constraints during the graph construction. Using the graph model, we mathematically formulate the problem and design an efficient algorithm to determine the transmission rate and coding strategy for eac...

  3. Application of data analysis techniques to nuclear reactor systems code to accuracy assessment

    International Nuclear Information System (INIS)

    An automated code assessment program (ACAP) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. This software was developed under subcontract to the United States Nuclear Regulatory Commission for use in its NRS code consolidation efforts. In this paper, background on the topic of NRS accuracy and uncertainty assessment is provided which motivates the development of and defines basic software requirements for ACAP. A survey of data analysis techniques was performed, focusing on the applicability of methods in the construction of NRS code-data comparison measures. The results of this review process, which further defined the scope, user interface and process for using ACAP are also summarized. A description of the software package and several sample applications to NRS data sets are provided. Its functionality and ability to provide objective accuracy assessment figures are demonstrated. (author)

  4. Classification and modelling of functional outputs of computation codes. Application to accidental thermal-hydraulic calculations in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    This research thesis has been made within the frame of a project on nuclear reactor vessel life. It deals with the use of numerical codes aimed at estimating probability densities for every input parameter in order to calculate probability margins at the output level. More precisely, it deals with codes with one-dimensional functional responses. The author studies the numerical simulation of a pressurized thermal shock on a nuclear reactor vessel, i.e. one of the possible accident types. The study of the vessel integrity relies on a thermal-hydraulic analysis and on a mechanical analysis. Algorithms are developed and proposed for each of them. Input-output data are classified using a clustering technique and a graph-based representation. A method for output dimension reduction is proposed, and a regression is applied between inputs and reduced representations. Applications are discussed in the case of modelling and sensitivity analysis for the CATHARE code (a code used at the CEA for the thermal-hydraulic analysis)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

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

  6. Fabrication and microstructural analysis of UN-U3Si2 composites for accident tolerant fuel applications

    Science.gov (United States)

    Johnson, Kyle D.; Raftery, Alicia M.; Lopes, Denise Adorno; Wallenius, Janne

    2016-08-01

    In this study, U3Si2 was synthesized via the use of arc-melting and mixed with UN powders, which together were sintered using the SPS method. The study revealed a number of interesting conclusions regarding the stability of the system - namely the formation of a probable but as yet unidentified ternary phase coupled with the reduction of the stoichiometry in the nitride phase - as well as some insights into the mechanics of the sintering process itself. By milling the silicide powders and reducing its particle size ratio compared to UN, it was possible to form a high density UN-U3Si2 composite, with desirable microstructural characteristics for accident tolerant fuel applications.

  7. Shipping container response to severe highway and railway accident conditions: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

  8. Shipping container response to severe highway and railway accident conditions: Appendices

    International Nuclear Information System (INIS)

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

  9. Application of GIS for population dose assessment in the Chernobyl accident area

    International Nuclear Information System (INIS)

    The development of updated approaches to evaluating the long-term consequences of the Chernobyl accident is based on the experience of authors in assessing population doses and in creating Decision Support Systems (DSS) in radioecology with the use of GIS technologies. 'Dose block' is a key component of the PRANA GIS-DSS on rehabilitation of contaminated territories after Chernobyl accident. When estimating internal and external doses to the local population the following components of PRANA are used: electronic maps for territories under investigation (5 contaminated districts of Bryansk reg.); databases (including database associated with polygons of vector electronic maps, and database for model and other input parameters); updated and modified mathematical models for assessing external and internal doses to the local population (from 137Cs and 90Sr); corresponding computer modules and user interface. Library of electronic maps includes different layers of vector maps of landuse for territories under consideration. Map of landuse for each district comprises all elements of land use: agricultural fields (arable lands, pastures and hayfields), forests, gardens, settlements, swamps and water bodies. Databases associated with polygons of vector maps include the following main monitoring data for each polygon: for fields - soil contamination density with 137Cs and 90Sr, soil type, mechanical and chemical composition, crop rotation; for settlements - contamination density, monitoring data of internal and external doses, local diet, behaviour and location and other factors, demographic data, etc. Other databases include all the parameters necessary for assessing contamination of agricultural products and internal/external doses: various transfer factors, parameters of countermeasures, productivity and/or production of agricultural crops, milk and meat (both for private and farm production), etc. Special attention was paid to developing and adjusting to elements of GIS

  10. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

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

    International Nuclear Information System (INIS)

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

  12. International assessment of application of the Code of Conduct on the Safety of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor

    2015-11-15

    The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; notably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

  13. APPLICATION OF TURBO CODES IN HIGH-SPEED REAL-TIME CHANNEL

    Institute of Scientific and Technical Information of China (English)

    Zhao Danfeng; Yue Li; Yang Jianhua

    2006-01-01

    The time delay of Turbo codes due to its iterative decoding is the main bottleneck of its application in real-time channel. However, the time delay can be greatly shortened through the adoption of parallel decoding algorithm, dividing the received bits into several sub-blocks and processing in parallel. This letter mainly discusses the applicability of turbo codes in high-speed real-time channel through the study of a parallel turbo decoding algorithm based on 3GPP-proposed turbo encoder and interleaver in various channel. Simulation result shows that, by choosing an appropriate sub-block length, the time delay can be obviously shortened without degrading the performance and increasing hardware complexity, and furthermore indicates the applicability of Turbo codes in high-speed real-time channel.

  14. International assessment of application of the Code of Conduct on the Safety of Research Reactors

    International Nuclear Information System (INIS)

    The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; notably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

  15. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  16. Accident management information needs

    International Nuclear Information System (INIS)

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs

  17. Accident and emergency management

    International Nuclear Information System (INIS)

    There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

  18. Plant application of ICARE/ASTECv2.0r3 computer code for investigation of in-vessel melt retention in VVER-1000 reactor design

    International Nuclear Information System (INIS)

    Highlights: • We have modeled external water cooling of the reactor vessel. • The NPP analyzed is a VVER1000 nuclear power plant. • ICARE/ASTECv2.0r3 stand-alone calculation has been done. • Heat fluxes from the corium to the vessel have been investigated. • Heat fluxes from the vessel to the water have been investigated. - Abstract: This investigation was done to assess the applicability of the In-Vessel Melt Retention (IVMR) strategy with external vessel water cooling for the reactors of VVER-1000/320 type. The IVMR strategy has been studied in different countries and the existing knowledge can assess the applicability of the strategy for VVER-1000/320 reactors and to identify the main remaining issues. IVMR strategy is one of the feasible solutions to mitigate severe accidents of VVER-1000/320 reactors but it needs to be further developed and optimized. The analytical work aimed at further identification and verification of conditions for successful IVMR application. The results achieved with severe accident computer codes like ASTEC could predict these conditions. The priority issues evaluated in this paper are the heat transfer from the corium to the reactor pressure vessel (RPV) and the heat transfer from the RPV to water outside the vessel. ICARE/ASTECv2.0r3 computer code was used to predict the heat fluxes from the corium to the vessel and the heat fluxes from the vessel to the outside coolant. It highlighted key periods of maximum heat input to the vessel steel wall that the steel wall has to be capable of sustaining

  19. Application and Implementation of Network Coding for Cooperative Wireless Networks

    DEFF Research Database (Denmark)

    Pedersen, Morten Videbæk

    2012-01-01

    Today the traditional client-server network architecture is the predominant model in our network infrastructure. However, for the increasing amount of \\live" services such as TV and radio being digitalized and the growing amount of user generated content, the centralized model can provide a poor...... utilization of the available network resources. To efficiently support these services we look towards the field of user cooperation. In order to create the incentive for users to join the cooperation we must make the gain larger than the expense. In this PhD I have suggested two central ways of achieving this...... ways that cooperative models may be implemented to cover a wide range of applications. This addresses the development of user cooperative protocols and how we in Device To Device (D2D) communication may reward users that contribute more to the network than they gain. In this area I suggest the use...

  20. An application of the LOCA code Cupidon: an assessment of the cladding behaviour in the flash tests

    International Nuclear Information System (INIS)

    The Cupidon code has been developed to analyze the thermo-mechanical behaviour of a fuel rod during a Loss Of Coolant Experiment. Models included are drawn from out-of-pile results such Edgar and the first use is to predict and calculate the tests carried out in the Phebus facility. Although the Flash program initiated at Grenoble is devoted to the study of fission product release during a LOCA (Loss Of Coolant Accident), interesting informations have been obtained on in-pile cladding deformation during transients. Analyses of the PIE (Post Irradiated Examination) results in the two first experiments with Cupidon code have shown fairly good agreement regarding diametral strain

  1. Recent advances and applications of the MAFIA codes

    Science.gov (United States)

    Wipf, S. G.; Marx, M.; Dohlus, M.; Steffen, B.; Blell, U.; Bartsch, M.; Hahne, P.; Schulz, A.; Schütt, P.; Wieland, T.; Becker, U.; Dehler, M.; Du, X.; Klatt, R.; Langstrof, A.; Pröpper, Zhang Min T.; van Rienen, U.; Schmitt, D.; Thoma, P.; Wagner, B.

    1993-12-01

    Over the last years MAFIA has grown to a more and more universal design tool for a vast range of applications not only in the field of accelerator physics. The currently distributed version 3.1 now includes a new solver module for time harmonic fields that enables the computation of eddy current distributions as well as the fields in driven rf systems. MAFIA 3.1 also includes static modules for electric and magnetic fields, 2D and 3D resonator solvers, 2D and 3D time domain solvers as well as 2.5D and 3D PIC modules. Thus MAFIA 3.1 now virtually covers the entire range of electromagnetic field problems. The fully menu driven user interface has been enhanced by implementation of macros, symbolic variables, and language structures that makes MAFIA fully programmable. On the application side there are numerous highlights such as extremely fast and accurate computations of S-parameters, calculation of antennas including farfield patterns, non-destructive testing analysis for carbon fiber reinforced plastic as used as air plane material, etc., to name only a few. In the accelerator physics area the new version added many enhancements on the calculation of impedances and wake fields with the possibility to simulate very short bunches without excessive need for memory. Version 3.2, scheduled for release in fall 1993, contains further new features such as fully lossy materials (complex fields), cylindrical coordinates for better cavity design, a possibility to add user-defined menus, various new 3D visualization tools, enhanced MAFIA language, and an AUTOMESH option. The most important new module is an optimizer, called OO, which basically combines all MAFIA modules into one (big) program. OO allows fully automatic optimization of electromagnetic components such as waveguide transitions, cavities, etc., according to user specified goal functions.

  2. Iterative ensemble Kalman filter for atmospheric dispersion in nuclear accidents: An application to Kincaid tracer experiment.

    Science.gov (United States)

    Zhang, X L; Su, G F; Chen, J G; Raskob, W; Yuan, H Y; Huang, Q Y

    2015-10-30

    Information about atmospheric dispersion of radionuclides is vitally important for planning effective countermeasures during nuclear accidents. Results of dispersion models have high spatial and temporal resolutions, but they are not accurate enough due to the uncertain source term and the errors in meteorological data. Environmental measurements are more reliable, but they are scarce and unable to give forecasts. In this study, our newly proposed iterative ensemble Kalman filter (EnKF) data assimilation scheme is used to combine model results and environmental measurements. The system is thoroughly validated against the observations in the Kincaid tracer experiment. The initial first-guess emissions are assumed to be six magnitudes underestimated. The iterative EnKF system rapidly corrects the errors in the emission rate and wind data, thereby significantly improving the model results (>80% reduction of the normalized mean square error, r=0.71). Sensitivity tests are conducted to investigate the influence of meteorological parameters. The results indicate that the system is sensitive to boundary layer height. When the heights from the numerical weather prediction model are used, only 62.5% of reconstructed emission rates are within a factor two of the actual emissions. This increases to 87.5% when the heights derived from the on-site observations are used.

  3. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    International Nuclear Information System (INIS)

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES

  4. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

  5. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in GE-designed operating plants and near-term operating license applications

    International Nuclear Information System (INIS)

    The results are presented of a generic evaluation of feedwater transients, small-break loss-of-coolant accidents (LOCAs), and other TMI-2-related events for General Electric Company (GE)-designed operating plants and near-term operating license applications to confirm or establish the bases for the continued safe operation of the operating plants. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas. Additional review of the accident is continuing and further information is being obtained and evaluated. Any new information will be reviewed and modifications will be made as appropriate

  6. Compiled reports on the applicability of selected codes and standards to advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

  7. Development of a shell finite element. Application to the thermo-viscoplastic behaviour of a PWR vessel during a severe accident

    International Nuclear Information System (INIS)

    The aim of this study is to develop a model for the thermo-viscoplastic behaviour of he power water reactor lower head during a severe accident, so as to implement it in codes representing the whole accident progress (scenario codes). So it has to give a precise solution in a short cpu-time. The main loadings are the internal pressure and the strong longitudinal and transverse thermal gradients. To deal with this problem, the idea is to develop a new shell element with variable mechanical parameters with the temperature. This is possible in taking advantage of the properties of the bending center line, called neutral fiber. Besides, this new shell element has the particularity to be able to melt without modifying the initial dimensions of the structure. Then, we have developed a complete program to study the mechanical resistance of the vessel. The visco-plastic behaviour is considered as a loading (so it is placed in the second member of the system to be solved) and represented by a Norton law whose parameters depend on the temperature, the law is integrated explicitly which necessitates the introduction of criteria limiting the time step. The rupture criterion by creep is defined by a damage law whereas the rupture criterion by plasticity is based on the exceeding of the mean limit stress in the thickness. Then the model was validated by comparing the results with those of a Castem 2000 volume mesh (finite element code). Finally the model was coupled with the scenario codes ICARE2 and MAAP4 and tested on two typical severe accidents. The results are very satisfactory both on accuracy and cpu-time execution. (author)

  8. Calculations of severe accident progression in the General Electric Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    General Electric is designing a new nuclear power plant: the Simplified Boiling Water Reactor (SBWR). The SBWR is a passive plant in which the core cooling and decay heat removal safety systems are driven by gravity. To model the plant response to severe accidents, MAAP-SBWR, an advanced version of the Modular Accident Analysis Program (MAAP), has been developed. The main feature of the new code is a flexible containment model. The challenges in modeling the SBWR, the code structure and models, and a sample application to the SBWR are discussed

  9. Analysis of hypothetical LMFBR whole-core accidents in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

    1978-01-01

    The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper.

  10. Validation and application of the MEWA 3D code for the investigation of multidimensional effects on debris coolability

    International Nuclear Information System (INIS)

    The coolability of particulate debris beds is a major issue during the course of a severe accident. In case that water is injected into the containment/reactor cavity as an accident mitigation measure there is the potential to stop the accident or at least mitigate its consequences if coolability of the debris bed can be achieved. Assessment of pros and cons of such measures requires sufficient knowledge and thus creates a challenge for the further development of severe accident codes. To give a more realistic description of the coolability during a severe accident, when molten corium is released into the reactor cavity and a debris bed is built up due to fragmentation of the ejected molten corium jet, a three dimensional code for MElt and WAter interaction, MEWA 3D, is being developed at developed at IKE, University of Stuttgart. MEWA 3D simulates the boil-off and quenching behavior of particulate or porous debris beds in multidimensional flow conditions. The aim is to describe the in- and ex-vessel behavior of solidified corium during the late phase of severe accidents in light water reactors, taking into account processes of core heat-up, melting, degradation and relocation either to the lower plenum or to the cavity. A brief description of the major modeling assumptions, governing conservation equations and constitutive laws used in the code will be given. The focus of this contribution is on the investigation of multidimensional aspects of the cooling behavior in order to analyze the coolability under reactor conditions. Taking into account the lateral pouring of the melt into the spreading room, the bed configuration most likely will be truly three-dimensional. For realistic simulations, it is essential to consider the real geometrical conditions, e.g. non-symmetrical debris configurations as result of the non-central core relocation. For the validation of the code for long-term coolability, MEWA 3D is applied to perform calculations of the COOLOCE

  11. Study on the distribution law of random code structure of irregular LDPC codes and its application in eliminating short cycles

    Institute of Scientific and Technical Information of China (English)

    MA LinHua; CHANG YiLin; Wang ShengDa

    2007-01-01

    The distribution law of the random code structure of randomly constructed irregular low-density parity-check (LDPC) codes is studied. Based on the Progressive Edge-Growth (PEG) algorithm, a new algorithm which can both eliminate short cycles and keep the distribution of the random code structure is presented. The experimental results show that the performance of the irregular LDPC codes constructed by the new algorithm is superior to that of the PEG algorithm.

  12. A Method for Modeling Co-Occurrence Propensity of Clinical Codes with Application to ICD-10-PCS Auto-Coding

    OpenAIRE

    Subotin, Michael; Davis, Anthony R.

    2015-01-01

    Objective. Natural language processing methods for medical auto-coding, or automatic generation of medical billing codes from electronic health records, generally assign each code independently of the others. They may thus assign codes for closely related procedures or diagnoses to the same document, even when they do not tend to occur together in practice, simply because the right choice can be difficult to infer from the clinical narrative. Materials and Methods. We propose a method that in...

  13. Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-05-01

    This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

  14. Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-04-01

    This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

  15. Application of bar codes to the automation of analytical sample data collection

    International Nuclear Information System (INIS)

    The Health Protection Department at the Savannah River Plant collects 500 urine samples per day for tritium analyses. Prior to automation, all sample information was compiled manually. Bar code technology was chosen for automating this program because it provides a more accurate, efficient, and inexpensive method for data entry. The system has three major functions: sample labeling is accomplished at remote bar code label stations composed of an Intermec 8220 (Intermec Corp.) interfaced to an IBM-PC, data collection is done on a central VAX 11/730 (Digital Equipment Corp.). Bar code readers are used to log-in samples to be analyzed on liquid scintillation counters. The VAX 11/730 processes the data and generates reports, data storage is on the VAX 11/730 and backed up on the plant's central computer. A brief description of several other bar code applications at the Savannah River Plant is also presented

  16. Application of ant colony optimization approach to severe accident management measures of Maanshan nuclear power plant

    International Nuclear Information System (INIS)

    The first three guidelines in the Maanshan SAMG were respectively evaluated for the effects in the SBO incident. The MAAP5 code was used to simulate the sequence of events and physical phenomena in the plant. The results show that the priority optimization should be carried out at two separated scenarios, i.e. the power recovered prior or after hot-leg creep rupture. The performance indices in the ant colony optimization could be the vessel life and the hydrogen generation from core for ant colony optimization. (author)

  17. Application of ant colony optimization approach to severe accident management measures of Maanshan nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, C.-M.; Wang, S.-J. [Inst. of Nuclear Energy Research, Taiwan (China)

    2011-07-01

    The first three guidelines in the Maanshan SAMG were respectively evaluated for the effects in the SBO incident. The MAAP5 code was used to simulate the sequence of events and physical phenomena in the plant. The results show that the priority optimization should be carried out at two separated scenarios, i.e. the power recovered prior or after hot-leg creep rupture. The performance indices in the ant colony optimization could be the vessel life and the hydrogen generation from core for ant colony optimization. (author)

  18. Development of the Approach by States method and thermodynamical study of a 1300 MWe PWR type reactor following a complete water loss of vapor generator alimentation with the Cathare 2 code; Developpement de la conduite APE et etude thermohydraulique d'un REP 1300 MWe suite a un accident de perte totale d'eau alimentaire des generateurs de vapeur avec le code Cathare 2

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, F

    1998-06-30

    The objective of this report is to study the thermohydraulic behavior of a 1300 MWe PWR type reactor for a complete loss accident in water supplying of vapor generators. The Cathare computer code has been used in this aim. (N.C.)

  19. Iterative ensemble Kalman filter for atmospheric dispersion in nuclear accidents: An application to Kincaid tracer experiment

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, X.L.; Su, G.F.; Chen, J.G. [Institute of Public Safety Research, Department of Engineering Physics, Tsinghua University, Beijing 100084 (China); Raskob, W. [Institute for Nuclear and Energy Technologies, Karlsruhe Institute of Technology, Karlsruhe, D-76021 (Germany); Yuan, H.Y., E-mail: hy-yuan@outlook.com [Institute of Public Safety Research, Department of Engineering Physics, Tsinghua University, Beijing 100084 (China); Huang, Q.Y. [Institute of Public Safety Research, Department of Engineering Physics, Tsinghua University, Beijing 100084 (China)

    2015-10-30

    Highlights: • We integrate the iterative EnKF method into the POLYPHEMUS platform. • We thoroughly evaluate the data assimilation system against the Kincaid dataset. • The data assimilation system substantially improves the model predictions. • More than 60% of the retrieved emissions are within a factor two of actual values. • The results reveal that the boundary layer height is the key influential factor. - Abstract: Information about atmospheric dispersion of radionuclides is vitally important for planning effective countermeasures during nuclear accidents. Results of dispersion models have high spatial and temporal resolutions, but they are not accurate enough due to the uncertain source term and the errors in meteorological data. Environmental measurements are more reliable, but they are scarce and unable to give forecasts. In this study, our newly proposed iterative ensemble Kalman filter (EnKF) data assimilation scheme is used to combine model results and environmental measurements. The system is thoroughly validated against the observations in the Kincaid tracer experiment. The initial first-guess emissions are assumed to be six magnitudes underestimated. The iterative EnKF system rapidly corrects the errors in the emission rate and wind data, thereby significantly improving the model results (>80% reduction of the normalized mean square error, r = 0.71). Sensitivity tests are conducted to investigate the influence of meteorological parameters. The results indicate that the system is sensitive to boundary layer height. When the heights from the numerical weather prediction model are used, only 62.5% of reconstructed emission rates are within a factor two of the actual emissions. This increases to 87.5% when the heights derived from the on-site observations are used.

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

    International Nuclear Information System (INIS)

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

  1. The Darlington emergency response projection code

    International Nuclear Information System (INIS)

    This paper describes the main features of the Darlington Emergency Response Projection code DERP, a self contained and user friendly personal computer program developed to supplement Ontario Hydro's emergency response capability fur nuclear power plants. DERP addresses the particular accident response characteristics of the negative pressure CANDU containment system, and makes dose projections for the area surrounding the Darlington Nuclear Generating Station resulting from airborne releases following a nuclear accident. Its main application is as an aid in the decision-making process regarding public protection strategies concerning off-site actions such as sheltering or evacuation

  2. Random wavelet transforms, algebraic geometric coding, and their applications in signal compression and de-noising

    Energy Technology Data Exchange (ETDEWEB)

    Bieleck, T.; Song, L.M.; Yau, S.S.T. [Univ. of Illinois, Chicago, IL (United States); Kwong, M.K. [Argonne National Lab., IL (United States). Mathematics and Computer Science Div.

    1995-07-01

    The concepts of random wavelet transforms and discrete random wavelet transforms are introduced. It is shown that these transforms can lead to simultaneous compression and de-noising of signals that have been corrupted with fractional noises. Potential applications of algebraic geometric coding theory to encode the ensuing data are also discussed.

  3. Applicability of small-scale integral test data to the 4500 MWt ESBWR loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Saha, Pradip [GE Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, NC 28401 (United States)], E-mail: pradip.saha@ge.com; Gamble, Robert E.; Shiralkar, Bharat S.; Fitch, James R. [GE Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, NC 28401 (United States)

    2009-05-15

    This paper discusses the scaling methodology used by GE Hitachi Nuclear Energy (GEH) to show that the data obtained from the small-scale integral test facilities, namely, GIST and GIRAFFE-SIT, are relevant to the postulated loss-of-coolant accident (LOCA) of the 4500 MWt ESBWR. The conservation of mass and energy equations for the steam-water mixture in the reactor pressure vessel (RPV) are transformed to the equations for the rates of pressure change and water mass or inventory change. These equations are non-dimensionalized based on the most dominant physical processes of the individual stages of a LOCA, namely, the late blowdown stage, the GDCS (gravity driven cooling system) transition stage and the full GDCS stage. The magnitudes of the non-dimensional Pi-groups, obtained from these equations, for the 4500 MWt ESBWR are compared with those obtained for the small-scale integral tests mentioned earlier. In addition, simplified analyses were conducted for the first two stages by integrating the non-dimensional RPV depressurization rate and the water inventory change rate equations. The results of the 4500 MWt ESBWR are very similar to the test data obtained from the GIST and the GIRAFFE-SIT test facilities. Therefore, based on both the Pi-group magnitudes and the simplified analyses, it is concluded that the small-scale integral test data mentioned above are applicable to the 4500 MWt ESBWR LOCA applications.

  4. Development of integrated transport code, TASK3D, and its applications to LHD experiment

    International Nuclear Information System (INIS)

    The integrated transport code for helical plasmas, TASK3D, has been developed both by modifying modules in TASK to be applicable to three-dimensional magnetic configurations, and by adding new modules for stellarator-heliotron specific physics and incorporating three-dimensional equilibria. In this paper, these module developments so far are collectively introduced, and recent progress on the applications of TASK3D to heat transport analyses of LHD plasmas is introduced. (author)

  5. The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

    Science.gov (United States)

    Iwamoto, O.; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

    2016-01-01

    A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

  6. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  7. THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

    International Nuclear Information System (INIS)

    THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

  8. Calculation of a Reactivity Initiated Accident with a 3D Cell-by-Cell Method: Application of the SAPHYR System to a Rod Ejection Accident in TMI1

    International Nuclear Information System (INIS)

    The 3D method for neutronic calculations of a Reactivity-Initiated Accident (RIA) in a Pressurized Water Reactor (PWR) tends now to replace the former 2D-1D evaluations. In the frame of a search of adequate penalties to apply to this new type of transient modeling, sensitivity studies on the geometrical description accuracy, on the type of physical phenomena modeled and on the values of the key physical parameters are currently undertaken by the IPSN and the CEA. 3D core neutronic calculations with a homogeneous and a heterogeneous (cell by cell) assembly modeling were compared. The influence of the axial neutronic calculation meshing was studied. As far as thermal-hydraulics is concerned, cross flow effects were measured by a comparison of 3D and 1D core FLICA4 models (each assembly being represented by one thermal-hydraulic channel). A more accurate feedback modeling was also studied by adjusting the thermal-hydraulic calculation meshes to the neutronic ones (four thermal-hydraulic channels per assembly). All these tests showed first that the power transient can be noticeably affected by the scale of the core physical modeling and by the type of physical phenomena taken into account (moderator feedback effects, core cross flows..). They also have highlighted that, for each studied model parameter, a preliminary search for the 'envelope transient' is necessary to be sure to evaluate its maximum impact on power transients. (authors)

  9. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  10. CowLog – Cross-Platform Application for Coding Behaviours from Video

    Directory of Open Access Journals (Sweden)

    Matti Pastell

    2016-04-01

    Full Text Available CowLog is a cross-platform application to code behaviours from video recordings for use in behavioural research. The software has been used in several studies e.g. to study sleep in dairy calves, emotions in goats and the mind wandering related to computer use during lectures. CowLog 3 is implemented using JavaScript and HTML using the Electron framework. The framework allows the development of packaged cross-platform applications using features from web browser (Chromium as well as server side JavaScript from Node.js. The program supports using multiple videos simultaneously and HTML5 and VLC video players. CowLog can be used for any project that requires coding the time of events from digital video. It is released under GNU GPL v2 making it possible for users to modify the application for their own needs. The software is available through its website http://cowlog.org.

  11. User's manual for the FEHM application - A finite-element heat- and mass-transfer code

    International Nuclear Information System (INIS)

    The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data, including individual input records or parameters, and the various output files. The system interface is described, including the software environment and installation instructions

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

    International Nuclear Information System (INIS)

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

  13. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rishel, Jeremy P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  14. Development and application of a deflagration pressure analysis code for high level waste processing

    International Nuclear Information System (INIS)

    The Deflagration Pressure Analysis Code (DPAC) was developed primarily to evaluate peak pressures for deflagrations in radioactive waste storage and process facilities at the Savannah River Site (SRS). Deflagrations in these facilities are generally considered to be incredible events, but it was judged prudent to develop modeling capabilities in order to facilitate risk estimates. DPAC is essentially an engineering analysis tool, as opposed to a detailed thermal hydraulics code. It accounts for mass loss via venting, energy dissipation by radiative heat transfer, and gas PdV work. Volume increases due to vessel deformation can also be included using pressure-volume data from a structural analysis of the enclosure. This paper presents an overview of the code, benchmarking, and applications at SRS

  15. Application of Bayesian Belief networks to the human reliability analysis of an oil tanker operation focusing on collision accidents

    International Nuclear Information System (INIS)

    During the last three decades, several techniques have been developed for the quantitative study of human reliability. In the 1980s, techniques were developed to model systems by means of binary trees, which did not allow for the representation of the context in which human actions occur. Thus, these techniques cannot model the representation of individuals, their interrelationships, and the dynamics of a system. These issues make the improvement of methods for Human Reliability Analysis (HRA) a pressing need. To eliminate or at least attenuate these limitations, some authors have proposed modeling systems using Bayesian Belief Networks (BBNs). The application of these tools is expected to address many of the deficiencies in current approaches to modeling human actions with binary trees. This paper presents a methodology based on BBN for analyzing human reliability and applies this method to the operation of an oil tanker, focusing on the risk of collision accidents. The obtained model was used to determine the most likely sequence of hazardous events and thus isolate critical activities in the operation of the ship to study Internal Factors (IFs), Skills, and Management and Organizational Factors (MOFs) that should receive more attention for risk reduction.

  16. RADTRAN: a computer code to analyze transportation of radioactive material

    International Nuclear Information System (INIS)

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

  17. Internal Accident Report on EDH

    CERN Multimedia

    SC Department

    2006-01-01

    The A2 Safety Code requires that, the Internal Accident Report form must be filled in by the person concerned or any witness to ensure that all the relevant services are informed. Please note that an electronic version of this form has been elaborated in collaboration with SC-IE, HR-OPS-OP and IT-AIS. Whenever possible, the electronic form shall be used. The relative icon is available on the EDH Desktop, Other tasks page, under the Safety heading, or directly here: https://edh.cern.ch/Document/Accident/. If you have any questions, please contact the SC Secretariat, tel. 75097 Please notice that the Internal Accident Report is an integral part of the Safety Code A2 and does not replace the HS50.

  18. Implementational Aspects of the Contourlet Filter Bank and Application in Image Coding

    Directory of Open Access Journals (Sweden)

    Truong T. Nguyen

    2009-02-01

    Full Text Available This paper analyzed the implementational aspects of the contourlet filter bank (or the pyramidal directional filter bank (PDFB, and considered its application in image coding. First, details of the binary tree-structured directional filter bank (DFB are presented, including a modification to minimize the phase delay factor and necessary steps for handling rectangular images. The PDFB is viewed as an overcomplete filter bank, and the directional filters are expressed in terms of polyphase components of the pyramidal filter bank and the conventional DFB. The aliasing effect of the conventional DFB and the Laplacian pyramid to the directional filters is then considered, and the conditions for reducing this effect are presented. The new filters obtained by redesigning the PDFBs satisfying these requirements have much better frequency responses. A hybrid multiscale filter bank consisting of the PDFB at higher scales and the traditional maximally decimated wavelet filter bank at lower scales is constructed to provide a sparse image representation. A novel embedded image coding system based on the image decomposition and a morphological dilation algorithm is then presented. The coding algorithm efficiently clusters the significant coefficients using progressive morphological operations. Context models for arithmetic coding are designed to exploit the intraband dependency and the correlation existing among the neighboring directional subbands. Experimental results show that the proposed coding algorithm outperforms the current state-of-the-art wavelet-based coders, such as JPEG2000, for images with directional features.

  19. Temporal Code-Driven Stimulation: Definition and Application to Electric Fish Signaling

    Science.gov (United States)

    Lareo, Angel; Forlim, Caroline G.; Pinto, Reynaldo D.; Varona, Pablo; Rodriguez, Francisco de Borja

    2016-01-01

    Closed-loop activity-dependent stimulation is a powerful methodology to assess information processing in biological systems. In this context, the development of novel protocols, their implementation in bioinformatics toolboxes and their application to different description levels open up a wide range of possibilities in the study of biological systems. We developed a methodology for studying biological signals representing them as temporal sequences of binary events. A specific sequence of these events (code) is chosen to deliver a predefined stimulation in a closed-loop manner. The response to this code-driven stimulation can be used to characterize the system. This methodology was implemented in a real time toolbox and tested in the context of electric fish signaling. We show that while there are codes that evoke a response that cannot be distinguished from a control recording without stimulation, other codes evoke a characteristic distinct response. We also compare the code-driven response to open-loop stimulation. The discussed experiments validate the proposed methodology and the software toolbox. PMID:27766078

  20. Second International Workshop on Software Engineering and Code Design in Parallel Meteorological and Oceanographic Applications

    Science.gov (United States)

    OKeefe, Matthew (Editor); Kerr, Christopher L. (Editor)

    1998-01-01

    This report contains the abstracts and technical papers from the Second International Workshop on Software Engineering and Code Design in Parallel Meteorological and Oceanographic Applications, held June 15-18, 1998, in Scottsdale, Arizona. The purpose of the workshop is to bring together software developers in meteorology and oceanography to discuss software engineering and code design issues for parallel architectures, including Massively Parallel Processors (MPP's), Parallel Vector Processors (PVP's), Symmetric Multi-Processors (SMP's), Distributed Shared Memory (DSM) multi-processors, and clusters. Issues to be discussed include: (1) code architectures for current parallel models, including basic data structures, storage allocation, variable naming conventions, coding rules and styles, i/o and pre/post-processing of data; (2) designing modular code; (3) load balancing and domain decomposition; (4) techniques that exploit parallelism efficiently yet hide the machine-related details from the programmer; (5) tools for making the programmer more productive; and (6) the proliferation of programming models (F--, OpenMP, MPI, and HPF).

  1. Generating Safety-Critical PLC Code From a High-Level Application Software Specification

    Science.gov (United States)

    2008-01-01

    The benefits of automatic-application code generation are widely accepted within the software engineering community. These benefits include raised abstraction level of application programming, shorter product development time, lower maintenance costs, and increased code quality and consistency. Surprisingly, code generation concepts have not yet found wide acceptance and use in the field of programmable logic controller (PLC) software development. Software engineers at Kennedy Space Center recognized the need for PLC code generation while developing the new ground checkout and launch processing system, called the Launch Control System (LCS). Engineers developed a process and a prototype software tool that automatically translates a high-level representation or specification of application software into ladder logic that executes on a PLC. All the computer hardware in the LCS is planned to be commercial off the shelf (COTS), including industrial controllers or PLCs that are connected to the sensors and end items out in the field. Most of the software in LCS is also planned to be COTS, with only small adapter software modules that must be developed in order to interface between the various COTS software products. A domain-specific language (DSL) is a programming language designed to perform tasks and to solve problems in a particular domain, such as ground processing of launch vehicles. The LCS engineers created a DSL for developing test sequences of ground checkout and launch operations of future launch vehicle and spacecraft elements, and they are developing a tabular specification format that uses the DSL keywords and functions familiar to the ground and flight system users. The tabular specification format, or tabular spec, allows most ground and flight system users to document how the application software is intended to function and requires little or no software programming knowledge or experience. A small sample from a prototype tabular spec application is

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

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

  4. Strain model and zircaloy-steam reaction model in the TUF code and their application in large LOCA

    International Nuclear Information System (INIS)

    As an integral part of a generic study of the Emergency Coolant Injection System effectiveness in Ontario Hydro reactors during a large break Loss of Coolant Accident (LOCA), the TUF (Two-Unequal-Fluids) code has been developed to enhance safety analysis capability. Recent enhancement to the TUF code includes the pressure tube transverse strain model and the zircaloy-steam reaction model. These models are employed to predict thermal-mechanical response of fuel channels and determine the thermal heat load to the moderator during postulated large LOCA scenarios. Presented in this paper are the description of the models, the cross-code comparison of the predictions between the TUF code and the SMARTT code and the discussion of parameters that may affect the pressure tube strain and the effect of pressure tube ballooning into contact with the calandria tube on the system response simulations. The modified TUF code is employed to quantify the extent of pressure tube ballooning and to calculate the thermal heat load to moderator. (author) 14 refs., 4 tabs., 16 figs

  5. Overview of HiFi -- implicit spectral element code framework for multi-fluid plasma applications

    CERN Document Server

    Lukin, Vyacheslav S; Lowrie, Weston; Meier, Eric T

    2016-01-01

    An overview of the algorithm and a sampling of plasma applications of the implicit, adaptive high order finite (spectral) element modeling framework, HiFi, is presented. The distinguishing capabilities of the HiFi code include adaptive spectral element spatial representation with flexible geometry, highly parallelizable implicit time advance, and general flux-source form of the partial differential equations and boundary conditions that can be implemented in its framework. Early algorithm development and extensive verification studies of the two-dimensional version of the code, known as SEL, have been previously described [A.H. Glasser & X.Z. Tang, Comp. Phys. Comm., 164 (2004); V.S. Lukin, Ph.D. thesis, Princeton University (2008)]. Here, substantial algorithmic improvements and extensions are presented together with examples of two- and three- dimensional applications of the HiFi framework. These include a Cartesian two-dimensional incompressible magnetohydrodynamic simulation of low dissipation magneti...

  6. Shared and Distributed Memory Parallel Security Analysis of Large-Scale Source Code and Binary Applications

    Energy Technology Data Exchange (ETDEWEB)

    Quinlan, D; Barany, G; Panas, T

    2007-08-30

    Many forms of security analysis on large scale applications can be substantially automated but the size and complexity can exceed the time and memory available on conventional desktop computers. Most commercial tools are understandably focused on such conventional desktop resources. This paper presents research work on the parallelization of security analysis of both source code and binaries within our Compass tool, which is implemented using the ROSE source-to-source open compiler infrastructure. We have focused on both shared and distributed memory parallelization of the evaluation of rules implemented as checkers for a wide range of secure programming rules, applicable to desktop machines, networks of workstations and dedicated clusters. While Compass as a tool focuses on source code analysis and reports violations of an extensible set of rules, the binary analysis work uses the exact same infrastructure but is less well developed into an equivalent final tool.

  7. The management of accidents

    Directory of Open Access Journals (Sweden)

    R. B. Ward

    2009-01-01

    Full Text Available Purpose: This author’s experiences in investigating well over a hundred accident occurrences has led to questioning how such events can be managed - - - while immediately recognising that the idea of managing accidents is an oxymoron, we don’t want to manage them, we don’t want not to manage them, what we desire is not to have to manage not-them, that is, manage matters so they don’t happen and then we don’t have to manage the consequences.Design/methodology/approach: The research will begin by defining some common classes of accidents in manufacturing industry, with examples taken from cases investigated, and by working backwards (too late, of course show how those involved could have managed these sample events so they didn’t happen, finishing with the question whether any of that can be applied to other situations.Findings: As shown that the management actions needed to prevent accidents are control of design and application of technology, and control and integration of people.Research limitations/implications: This paper has shown in some of the examples provided, management actions have been know to lead to accidents being committed by others, lower in the organization.Originality/value: Today’s management activities involve, generally, the use of technology in many forms, varying from simple tools (such as knives to the use of heavy equipment, electric power, and explosives. Against these we commit, in control of those items, the comparatively frail human mind and body, which, again generally, does succeed in controlling these resources, with (another generality by appropriate management. However, sometimes the control slips and an accident occurs.

  8. Tchernobyl accident

    International Nuclear Information System (INIS)

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given

  9. Infrared imaging :a proposed validation technique for computational fluid dynamics codes used in STOVL applications

    OpenAIRE

    Hardman, Robert R.

    1990-01-01

    The need for a validation technique for computational fluid dynamics (CFD) codes in STOVL applications has led to research efforts to apply infrared thermal imaging techniques to visualize gaseous flow fields. Specifically, a heated, free-jet test facility was constructed. The gaseous flow field of the jet exhaust was characterized using an infrared imaging technique in the 2 to 5.6μm wavelength band as well as conventional pitot tube and thermocouple methods. These infrared i...

  10. The Serpent Monte Carlo Code: Status, Development and Applications in 2013

    Science.gov (United States)

    Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

    2014-06-01

    The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

  11. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  12. 76 FR 30855 - Accident/Incident Reporting Requirements

    Science.gov (United States)

    2011-05-27

    ... accident/incident report (NTSB Recommendation Number RAR-05/02). See 75 FR 68891. However, upon further... NTSB Railroad Accident Report Number 05/02 (RAR 05/02). See 75 FR 68891. To clarify, FRA added Train... Equipment Accident/Incident Report.'' See 75 FR 68897. The codes represent the type of territory...

  13. A Protection Mechanism against Malicious HTML and JavaScript Code in Vulnerable Web Applications

    Directory of Open Access Journals (Sweden)

    Shukai Liu

    2016-01-01

    Full Text Available The high-profile attacks of malicious HTML and JavaScript code have seen a dramatic increase in both awareness and exploitation in recent years. Unfortunately, exiting security mechanisms provide no enough protection. We propose a new protection mechanism named PMHJ based on the support of both web applications and web browsers against malicious HTML and JavaScript code in vulnerable web applications. PMHJ prevents the injection attack of HTML elements with a random attribute value and the node-split attack by an attribute with the hash value of the HTML element. PMHJ ensures the content security in web pages by verifying HTML elements, confining the insecure HTML usages which can be exploited by attackers, and disabling the JavaScript APIs which may incur injection vulnerabilities. PMHJ provides a flexible way to rein the high-risk JavaScript APIs with powerful ability according to the principle of least authority. The PMHJ policy is easy to be deployed into real-world web applications. The test results show that PMHJ has little influence on the run time and code size of web pages.

  14. Bumpy Application of Utility Code for Genomic Inventions: With Special Reference to Express Sequence Tags

    Directory of Open Access Journals (Sweden)

    M R Sreenivasa Murthy

    2013-12-01

    Full Text Available Genomics, a new bough of biotechnology responsible for gene mapping has acquired a rapid significance in the field of patents. Brisk growth of patent filing in genomic subject matter is raising serious concerns about their utility from the perspective of societal benefit. Though the genomic related patent application qualifies the criterion of invention and non-obviousness in major instances, the inventors are unable to satisfy the utility criterion. Some instances such as patent application for ESTs have no utility at all. The patent regulators constructed various tests to deal with the situation such as specificity, substantiality (real world credibility tests etc. Hoverer, it is noteworthy that an attempt to uniform the standard of utility test for genomic inventions especially in the field of ESTs, cloning and creation of chimeras, has been made by America and Europe through specific regulations. Thus, the objective of this paper is firstly, to explain the importance of biotechnology and genomic inventions for mankind and significance of ESTs for future research. Secondly, to analyze the application of Utility code prior to the emergence of Utility code in America and Europe. Thirdly to scrutinize the Utility code in both countries and their implication on aftermath cases, and. fourthly and finally, to critically evaluate the both countries utility pathways in the light of societal benefit.

  15. A simplified Rietveld code for quantitative phase analysis: development, test and application to uranium mineral So

    International Nuclear Information System (INIS)

    As part of a team project about geological sampling in the environment of the city of Chihuahua, x-ray diffraction (XRD) phase analysis of different rock types is required. The most accepted technique to perform quantitative XRD phase analysis is the well-known Rietveld method. Rietveld codes (Full prof, Rietan, Rietica, DBWS, Topas,. . . ), oriented to a complete characterization of the diffraction pattern (from crystal structure to texture and crystal size investigation) have been developed by several authors. The majority of these codes show a high level of automation, but application may be long and troublesome anyway. False minima and instabilities during software running are recognized problems and represent current working lines of specialized groups. As the samples considered in the present geological investigation are particularly difficult for the Rietveld technique, it was decided to develop a program that performs phase analysis by an alternative route, with a degree of automation between Rietveld and the so-called Direct Comparison Method. The Basic Rietveld-ENhanced Diffraction Analysis (BRENDA) code has been developed and tested. BRENDA uses structure-factor (and other theoretical parameters) calculations from well-established diffraction codes (Full prof, Powder Cell), refines experimental diffraction peaks' profiles and intensities by means of a robust algorithm and determines phases' concentrations. Application of BRENDA code to computer-simulated problems, NIST standards and geological samples is divulged. Discrepancies with calibration figures are of the order of declared uncertainties. The advantages and disadvantages of the considered diffraction methods are discussed. In practice, being only half-automated gives the user more control of the refinement process and leads to an overall economy of time and higher reliability. (Author)

  16. Development of a shell finite element. Application to the thermo-viscoplastic behaviour of a PWR vessel during a severe accident; Developpement d`un element fini coque. Application au comportement thermo-viscoplastique d`une cuve de reacteur nucleaire (REP) en situation d`accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, V

    1998-10-07

    The aim of this study is to develop a model for the thermo-viscoplastic behaviour of he power water reactor lower head during a severe accident, so as to implement it in codes representing the whole accident progress (scenario codes). So it has to give a precise solution in a short cpu-time. The main loadings are the internal pressure and the strong longitudinal and transverse thermal gradients. To deal with this problem, the idea is to develop a new shell element with variable mechanical parameters with the temperature. This is possible in taking advantage of the properties of the bending center line, called neutral fiber. Besides, this new shell element has the particularity to be able to melt without modifying the initial dimensions of the structure. Then, we have developed a complete program to study the mechanical resistance of the vessel. The visco-plastic behaviour is considered as a loading (so it is placed in the second member of the system to be solved) and represented by a Norton law whose parameters depend on the temperature, the law is integrated explicitly which necessitates the introduction of criteria limiting the time step. The rupture criterion by creep is defined by a damage law whereas the rupture criterion by plasticity is based on the exceeding of the mean limit stress in the thickness. Then the model was validated by comparing the results with those of a Castem 2000 volume mesh (finite element code). Finally the model was coupled with the scenario codes ICARE2 and MAAP4 and tested on two typical severe accidents. The results are very satisfactory both on accuracy and cpu-time execution. (author) 113 refs.

  17. The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

    International Nuclear Information System (INIS)

    The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code

  18. A TSTT integrated FronTier code and its applications in computational fluid physics

    International Nuclear Information System (INIS)

    We introduce the FronTier-Lite software package and its adaptation to the TSTT geometry and mesh entity data interface. This package is extracted from the original front tracking code for general purpose scientific and engineering applications. The package contains a static interface library and a dynamic front propagation library. It can be used in research of different scientific problems. We demonstrate the application of FronTier in the simulations of fuel injection jet, the fusion pellet injection and fluid mixing problems

  19. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications

    International Nuclear Information System (INIS)

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs

  20. 14 CFR 415.41 - Accident investigation plan.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Accident investigation plan. 415.41 Section... Launch Range § 415.41 Accident investigation plan. An applicant must file an accident investigation plan... reporting and responding to launch accidents, launch incidents, or other mishaps, as defined by § 401.5...

  1. Light water reactor severe accident seminar. Seminar presentation manual

    International Nuclear Information System (INIS)

    The topics covered in this manual on LWR severe accidents were: Evolution of Source Term Definition and Analysis, Current Position on Severe Accident Phenomena, Current Position on Fission Product Behavior, Overview of Software Models Used in Severe Accident Analysis, Overview of Plant Specific Source Terms and Their Impact on Risk, Current Applications of Severe Accident Analysis, and Future plans

  2. A Methodology to Validate 3-D Arbitrary Lagrangian Eulerian Codes with Applications to Alegra

    Energy Technology Data Exchange (ETDEWEB)

    Chhabildas, L.C.; Duggins, B.D.; Konrad, C.H.; Mosher, D.A.; Perry, J.S.; Reinhart, W.D.; Summers, R.M.; Trucano, T.G.

    1998-11-04

    In this study we provided an experimental test bed for validating features of the Arbitrary Lagrangian Eulerian Grid for Research Applications (ALEGRA) code over a broad range of strain rates with overlapping diagnostics that encompass the multiple responses. A unique feature of the ALEGRA code is that it allows simultaneous computational treatment, within one code, of a wide range of strain-rates varying from hydrodynamic to structural conditions. This range encompasses strain rates characteristic of shock-wave propagation (107/s) and those characteristics of structural response (102/s). Most previous code validation experimental &udies, however, have been restricted to simulating or investigating a single strain-rate regime. What is new and different in this investigation is that we have performed well-controlled and well-instrumented experiments, which capture features relevant to both hydrodynamic and structural response in a single experiment. Aluminum was chosen for use in this study because it is a well-characterized material. The current experiments span strain rate regimes of over 107/s to less than 102/s in a single experiment. The input conditions were extremely well defined. Velocity interferometers were used to record the high' strain-rate response, while low strain rate data were collected using strain gauges. Although the current tests were conducted at a nominal velocity of - 1.5 km/s, it is the test methodology that is being emphasized herein. Results of a three-dimensional experiment are also presented.

  3. Application of the code Slac to the study of Ion Extraction Systems in Neutral Injectors

    International Nuclear Information System (INIS)

    In this study different extraction geometries for intense ion beams have been analyzed with the code SLAC, in view of its possible application to the neutral injectors of TJ-II. With this aim, we have introduced several modifications in the code in order to correctly simulate the transition between the ion source plasma and the extraction region, which has great impact on the beam optics. These modifications include the introduction of a population of Boltzmann electrons in the transition region, and the implementation of an option to simulate the thermal velocity of the ions in the source. We have found a better agreement between the results obtained with the new version of the code and the experimental data in two well known systems. With this new version of the code two different studies have been carried out: in the first place an optimization of the ATF injectors extraction system for its use on TJ-II, leading to an optimum value of the gap in the energy range 30-40 KeV, and in the second place a systematic study of extraction geometries at 40 KeV. As a result of this second study we have found the combinations of parameters that can be used under different working conditions (e.g. different pulse lengths), leading to acceptable values of the beam divergence. (Author)

  4. Tech-X Corporation releases simulation code for solving complex problems in plasma physics : VORPAL code provides a robust environment for simulating plasma processes in high-energy physics, IC fabrications and material processing applications

    CERN Multimedia

    2005-01-01

    Tech-X Corporation releases simulation code for solving complex problems in plasma physics : VORPAL code provides a robust environment for simulating plasma processes in high-energy physics, IC fabrications and material processing applications

  5. Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

    International Nuclear Information System (INIS)

    Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author)

  6. INF Code related matters. Joint IAEA/IMO literature survey on potential consequences of severe maritime accidents involving the transport of radioactive material. 2 volumes. Vol. I - Report and publication titles. Vol. II - Relevant abstracts

    International Nuclear Information System (INIS)

    This literature survey was undertaken jointly by the International Maritime Organization (IMO) and the International Atomic Energy Agency (IAEA) as a step in addressing the subject of environmental impact of accidents involving materials subject to the IMO's Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board Ships, also known as the INF Code. The results of the survey are provided in two volumes: the first one containing the description of the search and search results with the list of generated publication titles, and the second volume containing the abstracts of those publications deemed relevant for the purposes of the literature survey. Literature published between 1980 and mid-1999 was reviewed by two independent consultants who generated publication titles by performing searches of appropriate databases, and selected the abstracts of relevant publications for inclusion in this survey. The IAEA operates INIS, the world's leading computerised bibliographical information system on the peaceful uses of nuclear energy. The acronym INIS stands for International Nuclear Information System. INIS Members are responsible for determining the relevant nuclear literature produced within their borders or organizational confines, and then preparing the associated input in accordance with INIS rules. INIS records are included in other major databases such as the Energy, Science and Technology database of the DIALOG service. Because it is the INIS Members, rather than the IAEA Secretariat, who are responsible for its contents, it was considered appropriate that INIS be the primary source of information for this literature review. Selected unpublished reports were also reviewed, e.g. Draft Proceedings of the Special Consultative Meeting of Entities involved in the maritime transport of materials covered by the INF Code (SCM 5), March 1996. Many of the formal papers at SCM 5 were included in the literature

  7. A Novel Container ISO Code Localization Using an Object Clustering Method with Opencv and Visual Studio Application

    Directory of Open Access Journals (Sweden)

    Ronesh Sharma

    2013-06-01

    Full Text Available An automatic container code recognition system is of a great importance to the logistic supply chain management. Techniques have been proposed and implemented for the ISO container code region identification and recognition, however those systems have limitations on the type of container images with illumination factor and marks present on the container due to handling in the mass environmental condition. Moreover the research is not limited for differentiating between different formats of code and color of code characters. In this paper firstly an object clustering method is proposed to localize each line of the container code region. Secondly, the localizing algorithm is implemented with opencv and visual studio to perform localization and then recognition. Thus for real time application, the implemented system has added advantage of being easily integrated with other web application to increase the efficiency of the supply chain management. The experimental results and the application demonstrate the effectiveness of the proposed system for practical use.

  8. Transportation accidents

    International Nuclear Information System (INIS)

    Predicting the possible consequences of transportation accidents provides a severe challenge to an analyst who must make a judgment of the likely consequences of a release event at an unpredictable time and place. Since it is impractical to try to obtain detailed knowledge of the meteorology and terrain for every potential accident location on a route or to obtain accurate descriptions of population distributions or sensitive property to be protected (data which are more likely to be more readily available when one deals with fixed-site problems), he is constrained to make conservative assumptions in response to a demanding public audience. These conservative assumptions are frequently offset by very small source terms (relative to a fixed site) created when a transport vehicle is involved in an accident. For radioactive materials, which are the principal interest of the authors, only the most elementary models have been used for assessing the consequences of release of these materials in the transportation setting. Risk analysis and environmental impact statements frequently have used the Pasquill-Gifford/gaussian techniques for releases of short duration, which are both simple and easy to apply and require a minimum amount of detailed information. However, after deciding to use such a model, the problem of selecting what specific parameters to use in specific transportation situations still presents itself. Additional complications arise because source terms are not well characterized, release rates can be variable over short and long time periods, and mechanisms by which source aerosols become entrained in air are not always obvious. Some approaches that have been used to address these problems will be reviewed with emphasis on guidelines to avoid the Worst-Case Scenario Syndrome

  9. Utility implementation of EPRI rod ejection accident methodology

    International Nuclear Information System (INIS)

    This report describes the application of ARROTTA, a three dimensional space time kinetics code, to a licensing analysis of the PWR rod ejection accident. Three approaches for the use of ARROTTA are described: (1) a benchmark for point kinetics, (2) direct application as a biased licensing model, and (3) as a best estimate model used in conjunction with statistical combination of uncertainties. The use of ARROTTA as a biased licensing model was fully developed in conjunction with Duke Power Company; the results have been submitted to NRC as part of their reload licensing methodology

  10. Application of Gray Markov SCGM1,1c Model to Prediction of Accidents Deaths in Coal Mining

    OpenAIRE

    Lan, Jian-yi; Zhou, Ying

    2014-01-01

    The prediction of mine accident is the basis of aviation safety assessment and decision making. Gray prediction is suitable for such kinds of system objects with few data, short time, and little fluctuation, and Markov chain theory is just suitable for forecasting stochastic fluctuating dynamic process. Analyzing the coal mine accident human error cause, combining the advantages of both Gray prediction and Markov theory, an amended Gray Markov SCGM1,1c model is proposed. The gray SCGM1,1c mod...

  11. More efficient ground truth ROI image coding technique :implementation and wavelet based application analysis

    Institute of Scientific and Technical Information of China (English)

    KUMARAYAPA Ajith; ZHANG Ye

    2007-01-01

    In this paper, more efficient, low-complexity and reliable region of interest (ROI) image codec for compressing smooth low texture remote sensing images is proposed. We explore the efficiency of the modified ROI codec with respect to the selected set of convenient wavelet filters, which is a novel method. Such ROI coding experiment analysis representing low bit rate lossy to high quality lossless reconstruction with timing analysis is useful for improving remote sensing ground truth surveillance efficiency in terms of time and quality. The subjective [i.e. fair, five observer (HVS) evaluations using enhanced 3D picture view Hyper memory display technology] and the objective results revealed that for faster ground truth ROI coding applications, the Symlet-4 adaptation performs better than Biorthogonal 4.4 and Biorthogonal 6.8. However, the discrete Meyer wavelet adaptation is the best solution for delayed ROI image reconstructions.

  12. Tunable wavefront coded imaging system based on detachable phase mask: Mathematical analysis, optimization and underlying applications

    Science.gov (United States)

    Zhao, Hui; Wei, Jingxuan

    2014-09-01

    The key to the concept of tunable wavefront coding lies in detachable phase masks. Ojeda-Castaneda et al. (Progress in Electronics Research Symposium Proceedings, Cambridge, USA, July 5-8, 2010) described a typical design in which two components with cosinusoidal phase variation operate together to make defocus sensitivity tunable. The present study proposes an improved design and makes three contributions: (1) A mathematical derivation based on the stationary phase method explains why the detachable phase mask of Ojeda-Castaneda et al. tunes the defocus sensitivity. (2) The mathematical derivations show that the effective bandwidth wavefront coded imaging system is also tunable by making each component of the detachable phase mask move asymmetrically. An improved Fisher information-based optimization procedure was also designed to ascertain the optimal mask parameters corresponding to specific bandwidth. (3) Possible applications of the tunable bandwidth are demonstrated by simulated imaging.

  13. A Brain Computer Interface for Robust Wheelchair Control Application Based on Pseudorandom Code Modulated Visual Evoked Potential

    DEFF Research Database (Denmark)

    Mohebbi, Ali; Engelsholm, Signe K.D.; Puthusserypady, Sadasivan;

    2015-01-01

    In this pilot study, a novel and minimalistic Brain Computer Interface (BCI) based wheelchair control application was developed. The system was based on pseudorandom code modulated Visual Evoked Potentials (c-VEPs). The visual stimuli in the scheme were generated based on the Gold code...

  14. Melt spreading code assessment, modifications, and application to the EPR core catcher design

    International Nuclear Information System (INIS)

    The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted

  15. Melt spreading code assessment, modifications, and application to the EPR core catcher design.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T .; Nuclear Engineering Division

    2009-03-30

    The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial

  16. Application of Freeman Chain Codes: An Alternative Recognition Technique for Malaysian Car Plates

    CERN Document Server

    Jusoh, Nor Amizam

    2011-01-01

    Various applications of car plate recognition systems have been developed using various kinds of methods and techniques by researchers all over the world. The applications developed were only suitable for specific country due to its standard specification endorsed by the transport department of particular countries. The Road Transport Department of Malaysia also has endorsed a specification for car plates that includes the font and size of characters that must be followed by car owners. However, there are cases where this specification is not followed. Several applications have been developed in Malaysia to overcome this problem. However, there is still problem in achieving 100% recognition accuracy. This paper is mainly focused on conducting an experiment using chain codes technique to perform recognition for different types of fonts used in Malaysian car plates.

  17. Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

    International Nuclear Information System (INIS)

    In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result of this study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs

  18. Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Jung, Won Dae; Kang, Dae Il; Ha, Jae Joo

    1999-06-01

    In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result ofthis study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs.

  19. ICRP Publication 111 - Application of the Commission's recommendations to the protection of people living in long-term contaminated areas after a nuclear accident or a radiation emergency.

    Science.gov (United States)

    Lochard, J; Bogdevitch, I; Gallego, E; Hedemann-Jensen, P; McEwan, A; Nisbet, A; Oudiz, A; Oudiz, T; Strand, P; Janssens, A; Lazo, T; Carr, Z; Sugier, A; Burns, P; Carboneras, P; Cool, D; Cooper, J; Kai, M; Lecomte, J-F; Liu, H; Massera, G; McGarry, A; Mrabit, K; Mrabit, M; Sjöblom, K-L; Tsela, A; Weiss, W

    2009-06-01

    In this report, the Commission provides guidance for the protection of people living in long-term contaminated areas resulting from either a nuclear accident or a radiation emergency. The report considers the effects of such events on the affected population. This includes the pathways of human exposure, the types of exposed populations, and the characteristics of exposures. Although the focus is on radiation protection considerations, the report also recognises the complexity of post-accident situations, which cannot be managed without addressing all the affected domains of daily life, i.e. environmental, health, economic, social, psychological, cultural, ethical, political, etc. The report explains how the 2007 Recommendations apply to this type of existing exposure situation, including consideration of the justification and optimisation of protection strategies, and the introduction and application of a reference level to drive the optimisation process. The report also considers practical aspects of the implementation of protection strategies, both by authorities and the affected population. It emphasises the effectiveness of directly involving the affected population and local professionals in the management of the situation, and the responsibility of authorities at both national and local levels to create the conditions and provide the means favouring the involvement and empowerment of the population. The role of radiation monitoring, health surveillance, and the management of contaminated foodstuffs and other commodities is described in this perspective. The Annex summarises past experience of longterm contaminated areas resulting from radiation emergencies and nuclear accidents, including radiological criteria followed in carrying out remediation measures. PMID:20472181

  20. System-Level Genetic Codes Using a Transposable Element-Like Mechanism with Applications to Cancer

    OpenAIRE

    McGowan, John F.

    2000-01-01

    A system-level genetic code is a hypothetical genetic code that exclusively or preferentially codes systems of interacting coadapted parts. System-level genetic codes differ from part-level genetic codes in which each discrete part is coded independently. In general, a system-level genetic code requires coding discrete interacting parts such as organs or proteins in an interdependent way. Changing a single symbol or "gene" in a system-level genetic code affects two or more parts in a coordina...

  1. [Some consequences of the application of the new Swiss penal code on legal psychiatry].

    Science.gov (United States)

    Gasser, Jacques; Gravier, Bruno

    2007-09-19

    The new text of the Swiss penal code, which entered into effect at the beginning of 2007, has many incidences on the practice of the psychiatrists realizing expertises in the penal field or engaged in the application of legal measures imposing a treatment. The most notable consequences of this text are, on the one hand, a new definition of the concept of penal irresponsibility which is not necessarily any more related to a psychiatric diagnosis and, on the other hand, a new definition of legal constraints that justice can take to prevent new punishable acts and which appreciably modifies the place of the psychiatrists in the questions binding psychiatric care and social control.

  2. Scenario development on application of engineering technology for geological disposal. Study on engineering measures for accidents and human factors (Contract research)

    International Nuclear Information System (INIS)

    In the safety assessment for geological disposal of the high-level radioactive waste, scenarios need to be developed in consideration of influence on disposal systems by applying engineering technologies at each stage of site characterization, construction, operation and closure of disposal facility. To develop the scenarios, the engineering technologies which are applicable for each stage of geological disposal are listed in previous study. From this information, deviation events caused by the accidents and human factors lurking in the engineering technologies, which are deviated states of engineered and natural barriers from expected states occurred by applying engineering technologies were identified. Assuming the occurrence of the deviation events, possible evolution of features of barriers or loss/reduction of safety functions of barriers was discussed. Finally, the sequence of influence of the deviation events caused by application of engineering technologies on long-term safety after closure of the disposal facility was shown as scenarios. In this study, we compiled information of prevention measures for the accidents and human factors. Furthermore, we surveyed prevention measures and detecting means for the deviation events and compiled applicable influence reduction means for the deviation events. In addition, we identified remarkable deviation events from the point of view of safety, whose influence is not expected to be reduced sufficiently by these engineering measures. These results were integrated to the database that could support development of scenarios caused by application of engineering technologies to geological disposal. (author)

  3. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  4. Mathematical models for steam generator accident simulation

    International Nuclear Information System (INIS)

    In this contribution, the numerical methods used in the DeBeNe-LMFBR development for the analysis of the hydrodynamic and mechanical consequences of steam generator accidents are presented. At first the definition of the source term, i.e. the water leak rate which has to be assumed in the design basis accident as well as the thermochemistry of the sodium/water-reaction is discussed. Then the computer-codes presently used to describe the hydrodynamic and mechanical consequences of steam generator accidents on the basis of the above mentioned source term are presented. These comprise the code-system SAPHYR and the code PTANER and PISCES. Furthermore, developments which are planned or already under way for future use, such as the BEREPOT-code, are presented. (author)

  5. Scoping accident(s) for emergency planning

    International Nuclear Information System (INIS)

    At the request of the Conference of State Radiation Control Program Director's (CRCPD), in November 1976 the U.S. Nuclear Regulatory Commission formed a joint Task Force with representatives of the U.S. Environmental Protection Agency to answer a number of questions posed by the States regarding emergency planning. This Task Force held monthly meetings through November 1977. In December 1977 a draft report was prepared for limited distribution for review and comment by selected State and local organizations. The NRC/EPA Task Force deliberations centered on the CRCPD request for '... a determination of the most severe accident basis for which radiological emergency response plans should be developed by offsite agencies...' in the vicinity of nuclear power plants. Federal Interagency guidance to the States in this regard has been that the scoping accident should be the most serious conservatively analyzed accident considered for siting purposes, as exemplified in the Commission's Regulations at 10 CFR Part 100 and the NRC staffs Regulatory Guides 1.3 and 1.4, and as presented in license applicant's Safety Analysis Reports and the USNRC Staffs Safety Evaluation Reports. The draft report of the Task Force amplifies on this recommendation: to present a clearer picture of its import and introduces the concept of protective action zones (PAZs) within which detailed emergency plans should be developed; one zone for the plume exposure pathway and a second, larger zone for contamination pathways. The time dependence of potential releases and atmospheric transport, and important radionuclide groups of possible import are also discussed in the draft Task Force report. A status report regarding this effort, as of June 1978, will be presented. (author)

  6. International aspects of nuclear accidents

    International Nuclear Information System (INIS)

    The accident at Chernobyl revealed that there were shortcomings and gaps in the existing international mechanisms and brought home to governments the need for stronger measures to provide better protection against the risks of severe accidents. The main thrust of international co-operation with regard to nuclear safety issues is aimed at achieving a uniformly high level of safety in nuclear power plants through continuous exchanges of research findings and feedback from reactor operating experience. The second type of problem posed in the event of an accident resulting in radioactive contamination of several countries relates to the obligation to notify details of the circumstances and nature of the accident speedily so that the countries affected can take appropriate protective measures and, if necessary, organize mutual assistance. Giving the public accurate information is also an important aspect of managing an emergency situation arising from a severe accident. Finally, the confusion resulting from the unwarranted variety of protective measures implemented after the Chernobyl accident has highlighted the need for international harmonization of the principles and scientific criteria applicable to the protection of the public in the event of an accident and for a more consistent approach to emergency plans. The international conventions on third party liability in the nuclear energy sector (Paris/Brussels Conventions and the Vienna Convention) provide for compensation for damage caused by nuclear accidents in accordance with the rules and jurisdiction that they lay down. These provisions impose obligations on the operator responsible for an accident, and the State where the nuclear facility is located, towards the victims of damage caused in another country

  7. The FLUKA code for application of Monte Carlo methods to promote high precision ion beam therapy

    CERN Document Server

    Parodi, K; Cerutti, F; Ferrari, A; Mairani, A; Paganetti, H; Sommerer, F

    2010-01-01

    Monte Carlo (MC) methods are increasingly being utilized to support several aspects of commissioning and clinical operation of ion beam therapy facilities. In this contribution two emerging areas of MC applications are outlined. The value of MC modeling to promote accurate treatment planning is addressed via examples of application of the FLUKA code to proton and carbon ion therapy at the Heidelberg Ion Beam Therapy Center in Heidelberg, Germany, and at the Proton Therapy Center of Massachusetts General Hospital (MGH) Boston, USA. These include generation of basic data for input into the treatment planning system (TPS) and validation of the TPS analytical pencil-beam dose computations. Moreover, we review the implementation of PET/CT (Positron-Emission-Tomography / Computed- Tomography) imaging for in-vivo verification of proton therapy at MGH. Here, MC is used to calculate irradiation-induced positron-emitter production in tissue for comparison with the +-activity measurement in order to infer indirect infor...

  8. Energy Scaling Advantages of Resistive Memory Crossbar Based Computation and its Application to Sparse Coding

    Directory of Open Access Journals (Sweden)

    Sapan eAgarwal

    2016-01-01

    Full Text Available The exponential increase in data over the last decade presents a significant challenge to analytics efforts that seek to process and interpret such data for various applications. Neural-inspired computing approaches are being developed in order to leverage the computational advantages of the analog, low-power data processing observed in biological systems. Analog resistive memory crossbars can perform a parallel read or a vector-matrix multiplication as well as a parallel write or a rank-1 update with high computational efficiency. For an NxN crossbar, these two kernels are at a minimum O(N more energy efficient than a digital memory-based architecture. If the read operation is noise limited, the energy to read a column can be independent of the crossbar size (O(1. These two kernels form the basis of many neuromorphic algorithms such as image, text, and speech recognition. For instance, these kernels can be applied to a neural sparse coding algorithm to give an O(N reduction in energy for the entire algorithm. Sparse coding is a rich problem with a host of applications including computer vision, object tracking, and more generally unsupervised learning.

  9. Energy Scaling Advantages of Resistive Memory Crossbar Based Computation and Its Application to Sparse Coding.

    Science.gov (United States)

    Agarwal, Sapan; Quach, Tu-Thach; Parekh, Ojas; Hsia, Alexander H; DeBenedictis, Erik P; James, Conrad D; Marinella, Matthew J; Aimone, James B

    2015-01-01

    The exponential increase in data over the last decade presents a significant challenge to analytics efforts that seek to process and interpret such data for various applications. Neural-inspired computing approaches are being developed in order to leverage the computational properties of the analog, low-power data processing observed in biological systems. Analog resistive memory crossbars can perform a parallel read or a vector-matrix multiplication as well as a parallel write or a rank-1 update with high computational efficiency. For an N × N crossbar, these two kernels can be O(N) more energy efficient than a conventional digital memory-based architecture. If the read operation is noise limited, the energy to read a column can be independent of the crossbar size (O(1)). These two kernels form the basis of many neuromorphic algorithms such as image, text, and speech recognition. For instance, these kernels can be applied to a neural sparse coding algorithm to give an O(N) reduction in energy for the entire algorithm when run with finite precision. Sparse coding is a rich problem with a host of applications including computer vision, object tracking, and more generally unsupervised learning. PMID:26778946

  10. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  11. Systematic approach for assessment of accident risks in chemical and nuclear processing

    International Nuclear Information System (INIS)

    The industrial accidents which occurred in the last years, particularly in the 80's, contributed a significant way to draw the attention of the government, industry and the society as a whole to the mechanisms for preventing events that could affect people's safety and the environment quality. Techniques and methods extensively used the nuclear, aeronautic and war industries so far were adapted to performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. The risk analysis in industrial facilities is carried out through the evaluation of the probability or frequency of the accidents and their consequences. However, no systematized methodology that could supply the tools for identifying possible accidents likely to take place in an installation is available in the literature. Neither existing are methodologies for the identification of the models for evaluation of the accidents' consequences nor for the selection of the available techniques for qualitative or quantitative analysis of the possibility of occurrence of the accident being focused. The objective of this work is to develop and implement a methodology for identification of the risks of accidents in chemical and nuclear processing facilities as well as for the evaluation of their consequences on persons. For the development of the methodology, the main possible accidents that could occur in such installations were identified and the qualitative and quantitative techniques available for the identification of the risks and for the evaluation of the consequences of each identified accidents were selected. The use of the methodology was illustrated by applying it in two case examples adapted from the literature, involving accidents with inflammable, explosives, and radioactive materials. The computer code MRA - Methodology for Risk Assessment was developed using DELPHI, version 5.0, with the purpose of systematizing

  12. Contribution of Case Based Reasoning (CBR) in the Exploitation of Return of Experience. Application to Accident Scenarii in Railroad Transport

    CERN Document Server

    Maalel, Ahmed

    2012-01-01

    The study is from a base of accident scenarii in rail transport (feedback) in order to develop a tool to share build and sustain knowledge and safety and secondly to exploit the knowledge stored to prevent the reproduction of accidents / incidents. This tool should ultimately lead to the proposal of prevention and protection measures to minimize the risk level of a new transport system and thus to improve safety. The approach to achieving this goal largely depends on the use of artificial intelligence techniques and rarely the use of a method of automatic learning in order to develop a feasibility model of a software tool based on case based reasoning (CBR) to exploit stored knowledge in order to create know-how that can help stimulate domain experts in the task of analysis, evaluation and certification of a new system.

  13. Development and application of a random walk model of atmospheric diffusion in the emergency response of nuclear accidents

    Institute of Scientific and Technical Information of China (English)

    CHI Bing; LI Hong; FANG Dong

    2007-01-01

    Plume concentration prediction is one of the main contents of radioactive consequence assessment for early emergency response to nuclear accidents. Random characteristics of atmospheric diffusion itself was described, a random walk model of atmospheric diffusion (Random Walk) was introduced and compared with the Lagrangian puff model (RIMPUFF) in the nuclear emergency decision support system (RODOS) developed by the European Community for verification. The results show the concentrations calculated by the two models are quite close except that the plume area calculated by Random Walk is a little smaller than that by RIMPUFF. The random walk model for atmospheric diffusion can simulate the atmospheric diffusion in case of nuclear accidents, and provide more actual information for early emergency and consequence assessment as one of the atmospheric diffusion module of the nuclear emergency decision support system.

  14. Evolution of developments and applications of advanced thermal-hydraulics and neutronic codes. Conclusions from Annapolis Workshop and Ankara Seminar, Objectives of the Present Workshop

    International Nuclear Information System (INIS)

    In the nuclear reactor safety area, during the last 30-40 years, thermal-hydraulics has been one of the key disciplines for simulation and analysis of transient and accident scenarios and also for the definition of preventive and mitigative measures in relation to these scenarios. A workshop was organised by OECD/NEA-CSNI at Annapolis (1996) where codes, physical models, numeric and new computer architecture were examined. In parallel a Specialist meeting on instrumentation in two phase flows was held in Santa Barbara beginning of 1997 in order to investigate new techniques for getting measurements of new physical parameters necessary for assessing the new physical models. Among the different applications of thermal-hydraulic codes, the use of Best Estimate methods in safety evaluation is certainly one of the major challenges for which the safety and economic issues are quite important. For these reasons OECD/NEA-CSNI organised a seminar in Ankara in 1998 entirely devoted to the use of Best Estimate methods in thermal-hydraulics analysis. This seminar allowed to get a better view of where we were in such applications and which were the remaining problems and issues. The present workshop held in Barcelona beginning of year 2000 will be a good opportunity for providing an updated review of the gained progresses and for analysing if the objectives and programs are still progressing in the right direction. In order to do such exercise, we will first recall the questions which were raised in Annapolis and the main conclusions which were drawn from these questions. The conclusions of Ankara Meeting will be reviewed in a second step. Finally we will list the objectives of this workshop in Barcelona which is held in the continuity of Annapolis Workshop and Ankara Seminar. (authors)

  15. Application of a step-by-step fingerprinting identification method on a spilled oil accident in the Bohai Sea area

    Science.gov (United States)

    Sun, Peiyan; Gao, Zhenhui; Cao, Lixin; Wang, Xinping; Zhou, Qing; Zhao, Yuhui; Li, Guangmei

    2011-03-01

    In recent years, oil spill accidents occur frequently in the marine area of China. Finding out the spilled oil source is a key step in the relevant investigation. In this paper, a step-by-step fingerprinting identification method was used in a spilled oil accident in the Bohai Sea in 2002. Advanced chemical fingerprinting and data interpretation techniques were used to characterize the chemical composition and determine the possible sources of two spilled oil samples. The original gas chromatography -flame ionization detection (GC-FID) chromatogram of saturated hydrocarbons was compared. The gas chromatography-mass spectrometry (GC/MS) chromatograms of aromatic hydrocarbons terpane and sterane, n-alkane and poly-aromatic hydrocarbons (PAHs) were analyzed. The correlation analysis on diagnostic ratios was performed with Student's t-test. It is found that the oil fingerprinting of the spilled oil (designated as sz1) from the polluted sand beach was identical with the suspected oil (designated as ky1) from a nearby crude oil refinery factory. They both showed the fingerprinting character of mixed oil. The oil fingerprinting of the spilled oil (designated as ms1) collected from the port was significantly different from oil ky1 and oil sz1 and was with a lubricating oil fingerprint character. The identification result not only gave support for the spilled oil investigation, but also served as an example for studying spilled oil accidents.

  16. Application of a Step-by-Step Fingerprinting Identification Method on a Spilled Oil Accident in the Bohai Sea Area

    Institute of Scientific and Technical Information of China (English)

    SUN Peiyan; GAO Zhenhui; CAO Lixin; WANG Xinping; ZHOU Qing; ZHAO Yuhui; LI Guangmei

    2011-01-01

    In recent years, oil spill accidents occur frequently in the marine area of China. Finding out the spilled oil source is a key step in the relevant investigation. In this paper, a step-by-step fingerprinting identification method was used in a spilled oil accident in the Bohai Sea in 2002. Advanced chemical fingerprinting and data interpretation techniques were used to characterize the chemical composition and determine the possible sources of two spilled oil samples. The original gas chromatography -flame ionization detection (GC-FID) chromatogram of saturated hydrocarbons was compared. The gas chromatography-mass spectrometry (GC/MS)chromatograms of aromatic hydrocarbons terpane and sterane, n-alkane and poly-aromatic hydrocarbons (PAHs) were analyzed. The correlation analysis on diagnostic ratios was performed with Student's t-test. It is found that the oil fingerprinting of the spilled oil (designated as szl) from the polluted sand beach was identical with the suspected oil (designated as kyl) from a nearby crude oil refinery factory. They both showed the fingerprinting character of mixed oil. The oil fingerprinting of the spilled oil (designated as msl) collected from the port was significantly different from oil kyl and oil szl and was with a lubricating oil fingerprint character. The identification result not only gave support for the spilled oil investigation, but also served as an example for studying spilled oil accidents.

  17. Derived intervention levels for application in controlling radiation doses to the public in the event of a nuclear accident or radiological emergency

    International Nuclear Information System (INIS)

    This document sets out the principles and procedures for estimating derived intervention levels (DILs) and illustrates the application of these procedures to the estimation of DILs for a range of nuclides and exposure pathways, as aids to decisions on the application of protective measures for the public in the early and intermediate phases of a nuclear accident. The levels are derived subject to a number of assumptions about the intervention level of dose, the characteristics of the released material, the habits of the exposed individuals and local environmental conditions. Some guidance is given on the sensitivity of the estimated DILs to plausible variations in the above assumptions. The detailed procedures described in the document for estimating DILs and illustrations of their application are limited to accidental releases to the atmosphere

  18. Facilitating Code Reuse for the Rapid Deployment of Web Mapping Applications at the National Renewable Energy Laboratory (NREL)

    Science.gov (United States)

    Helm, C. W.; Sparks, W.; Levene, J.; Hostetler, M.

    2008-12-01

    The National Renewable Energy Laboratory in Golden, CO has developed a software platform that provides for the development of fully customized and unique web mapping applications that reuse a common base of software code. The application capabilities that have been developed within this platform include spatial data visualization, large-scale data retrieval and the analysis of various renewable energy resource data-sets. The platform consists of three primary components of reusable code: the back-end data storage and retrieval engine, a user-customizable Data Styling Engine, and front end user interface code. Each component of the platform represents a reusable code base from which new applications can be generated with a minimal amount of new code. This reusable code base can be thought of in the same vein as object oriented development: the reusable code is analogous to a base class that specific applications inherit from and extend. The architecture was motivated by a requirement to rapidly develop and deploy multiple web-based mapping applications for varying renewable energy and alternative fuel technologies, and for different customers. It was observed that these applications share a significant set of core features and functionality, with varying degrees of customization required for each application. A series of needs instigated the development of the architecture: * New applications should not require re-implementation of existing functionality (either through re-coding or "copy and paste" reuse) * Enhancements to the base functionality could automatically propagate through all derived applications * All applications should be able to utilize a common, internal (to NREL) Web Mapping Service (WMS), or any external WMS * The framework must support user authentication, role-based access control to specific data layers, and user customization of layer styling. This requirement led to the development of the Data Styling Engine. * A developer should be able to

  19. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    For the assessment of the safety of nuclear power plants it is of great importance the analyses of severe accidents since they allow to estimate the possible failure models of the containment, and also permit knowing the magnitude and composition of the radioactive material that would be released to the environment in case of an accident upon population and the environment. This paper presents in general terms the basic principles for conducting the analysis of severe accidents, the fundamental sources in the generation of radionuclides and aerosols, the transportation and deposition processes, and also makes reference to de main codes used in the modulation of severe accidents. The final part of the paper contents information on how severe accidents are dialed with the regulatory point view in different countries

  20. The Fukushima Daiichi Accident Study Information Portal

    Energy Technology Data Exchange (ETDEWEB)

    Shawn St. Germain; Curtis Smith; David Schwieder; Cherie Phelan

    2012-11-01

    This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

  1. Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application%ATHLET程序的钠冷快堆应用扩展及其验证

    Institute of Scientific and Technical Information of China (English)

    周翀; Klaus Huber; 程旭

    2013-01-01

    System analysis code is important for the global simulation of the sodium-cooled fast reactor (SFR) system as well as transient and accident safety analysis .In this paper ,the best estimate system code ATHLET for light water reactors ,developed by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) in Germany ,was modified for SFR application .Thermal-dynamic and transport properties as well as heat transfer correlations for sodium were implemented into the ATHLET code .The modified code was then applied to simulate the Phenix reactor in France ,and validation of the code was conducted with the Phenix reactor natural convection test .The calculation results were compared with the test data .The results show that the modified ATHLET code has good applicability in simulating SFR systems .%系统分析程序是对钠冷快堆的冷却剂回路系统进行全局模拟、瞬态及事故安全分析的重要工具。本工作对德国核设施与反应堆安全机构(GRS)开发的轻水堆最佳估算系统程序ATHLET 进行修改,增加了钠的物性公式和传热关系式,将其适用范围扩展到钠冷快堆。为验证修改过的ATHLET程序,对法国凤凰(Phenix )反应堆系统建模,并对其自然对流实验进行模拟,将计算结果与实验数据进行比较。结果显示,ATHLET程序的钠冷快堆应用扩展具有良好的适用性。

  2. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    International Nuclear Information System (INIS)

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  3. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  4. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

  6. Successive Wyner-Ziv Coding Scheme and its Application to the Quadratic Gaussian CEO Problem

    OpenAIRE

    Chen, Jun; Berger, Toby

    2006-01-01

    We introduce a distributed source coding scheme called successive Wyner-Ziv coding. We show that any point in the rate region of the quadratic Gaussian CEO problem can be achieved via the successive Wyner-Ziv coding. The concept of successive refinement in the single source coding is generalized to the distributed source coding scenario, which we refer to as distributed successive refinement. For the quadratic Gaussian CEO problem, we establish a necessary and sufficient condition for distrib...

  7. Application of the source term code package to obtain a specific source term for the Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    The main objective of the project was to use the Source Term Code Package (STCP) to obtain a specific source term for those accident sequences deemed dominant as a result of probabilistic safety analyses (PSA) for the Laguna Verde Nuclear Power Plant (CNLV). The following programme has been carried out to meet this objective: (a) implementation of the STCP, (b) acquisition of specific data for CNLV to execute the STCP, and (c) calculations of specific source terms for accident sequences at CNLV. The STCP has been implemented and validated on CDC 170/815 and CDC 180/860 main frames as well as on a Micro VAX 3800 system. In order to get a plant-specific source term, data on the CNLV including initial core inventory, burn-up, primary containment structures, and materials used for the calculations have been obtained. Because STCP does not explicitly model containment failure, dry well failure in the form of a catastrophic rupture has been assumed. One of the most significant sequences from the point of view of possible off-site risk is the loss of off-site power with failure of the diesel generators and simultaneous loss of high pressure core spray and reactor core isolation cooling systems. The probability for that event is approximately 4.5 x 10-6. This sequence has been analysed in detail and the release fractions of radioisotope groups are given in the full report. 18 refs, 4 figs, 3 tabs

  8. Application of tritium behavior simulation code (TBEHAVIOR) to an actual-scale tritium handling room

    International Nuclear Information System (INIS)

    It is essential from the viewpoint of fusion safety to confine and remove tritium in a room since tritium handling room is placed as 'final barrier' of fusion plant to prevent the environmental discharge of tritium. At the Tritium Process Laboratory (TPL) of Japan Atomic Energy Agency (JAEA), the application of our original three-dimensional TBEHAVIOR code to the tritium behavior in a room of 3000 m3 was verified. The Renormalization Group Theory (RNG) model was selected as Low-Reynolds model for practical calculation time as well as to reasonable precision in evaluation of velocity from the engineering viewpoint. A series of evaluated results indicated that a flow adjacent to a wall surface plays an important role for tritium transport in a ventilated room. Evaluation of attenuating behavior is further important since the ventilation is normally stopped for the tritium confinement in the case of tritium leakage. We demonstrated that an attenuating behavior can also be evaluated well by the TBEHAVIOR code. Even an attenuating or stagnant flow of less than 10mm/s in a room mixed tritium concentration uniform promptly. The presence of apparatuses in a room did not generally affect tritium behavior. Although the effect of buoyancy was limited to the initial period after the leak, the spread of tritium was promoted by buoyancy. It led to the shortening of elapsed time until the concentration became uniform. (author)

  9. Novel Polynomial Basis with Fast Fourier Transform and Its Application to Reed-Solomon Erasure Codes

    KAUST Repository

    Lin, Sian-Jheng

    2016-09-13

    In this paper, we present a fast Fourier transform (FFT) algorithm over extension binary fields, where the polynomial is represented in a non-standard basis. The proposed Fourier-like transform requires O(h lg(h)) field operations, where h is the number of evaluation points. Based on the proposed Fourier-like algorithm, we then develop the encoding/ decoding algorithms for (n = 2m; k) Reed-Solomon erasure codes. The proposed encoding/erasure decoding algorithm requires O(n lg(n)), in both additive and multiplicative complexities. As the complexity leading factor is small, the proposed algorithms are advantageous in practical applications. Finally, the approaches to convert the basis between the monomial basis and the new basis are proposed.

  10. Application of a Monte Carlo Penelope code at diverse dosimetric problems in radiotherapy

    International Nuclear Information System (INIS)

    In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: 192 Ir, 125 I, 106 Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)

  11. A portable platform for accelerated PIC codes and its application to GPUs using OpenACC

    CERN Document Server

    Hariri, F; Jocksch, A; Lanti, E; Progsch, J; Messmer, P; Brunner, S; Gheller, G; Villard, L

    2016-01-01

    We present a portable platform, called PIC_ENGINE, for accelerating Particle-In-Cell (PIC) codes on heterogeneous many-core architectures such as Graphic Processing Units (GPUs). The aim of this development is efficient simulations on future exascale systems by allowing different parallelization strategies depending on the application problem and the specific architecture. To this end, this platform contains the basic steps of the PIC algorithm and has been designed as a test bed for different algorithmic options and data structures. Among the architectures that this engine can explore, particular attention is given here to systems equipped with GPUs. The study demonstrates that our portable PIC implementation based on the OpenACC programming model can achieve performance closely matching theoretical predictions. Using the Cray XC30 system, Piz Daint, at the Swiss National Supercomputing Centre (CSCS), we show that PIC_ENGINE running on an NVIDIA Kepler K20X GPU can outperform the one on an Intel Sandybridge ...

  12. [Series: Medical Applications of the PHITS Code (2): Acceleration by Parallel Computing].

    Science.gov (United States)

    Furuta, Takuya; Sato, Tatsuhiko

    2015-01-01

    Time-consuming Monte Carlo dose calculation becomes feasible owing to the development of computer technology. However, the recent development is due to emergence of the multi-core high performance computers. Therefore, parallel computing becomes a key to achieve good performance of software programs. A Monte Carlo simulation code PHITS contains two parallel computing functions, the distributed-memory parallelization using protocols of message passing interface (MPI) and the shared-memory parallelization using open multi-processing (OpenMP) directives. Users can choose the two functions according to their needs. This paper gives the explanation of the two functions with their advantages and disadvantages. Some test applications are also provided to show their performance using a typical multi-core high performance workstation.

  13. Accident prevention programme

    International Nuclear Information System (INIS)

    This study by the Steel Industry Safety and Health Commission was made within the context of the application by undertakings of the principles of accident and disease prevention previously adopted by the said Commission. It puts forward recommendations for the effective and gradual implementation of a programme of action on occupational health and safety in the various departments of an undertaking and in the undertaking as a whole. The methods proposed in this study are likely to be of interest to all undertakings in the metallurgical industry and other industrial sectors

  14. Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

    Energy Technology Data Exchange (ETDEWEB)

    Arndt, S.A.

    1997-07-01

    The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.

  15. Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

    International Nuclear Information System (INIS)

    The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities

  16. Harmonic analysis of occupational-accident time-series as a part of the quantified risk evaluation in worksites: Application on electric power industry and construction sector

    International Nuclear Information System (INIS)

    The development of an integrated risk analysis scheme, which will combine a well-considered selection of widespread techniques, would enable the companies to achieve efficient results on risk assessment. In this study, we develop a methodological framework (as a part of the quantified risk evaluation), by incorporating a new technique, that is implemented by the harmonic-analysis of time-series of occupational-accidents (called as HATS). Our objective is therefore, twofold: (i) the development of a new risk assessment framework (HATS technique) and the subsequent application of HATS on the worksites of electric power industry and construction sector, and (ii) the enrichment of the harmonic-analysis theoretical background, as far as the significance-level of spectral peaks is concerned, with fully-completed practical tables, that they have been produced by using the scientific literature. In fact, we apply HATS on occupational-accident time-series, which were (a) observed in the worksites of the Greek Public electric Power Corporation (PPC) and the Greek construction-companies (GCCs), and (b) recorded in great statistical-databases of PPC, and IKA (the Greek Social Insurance Institute/Ministry of Health), respectively. The results of HATS were tested statistically by using Shimshoni's significance-test. Moreover, the results of the comparative time/frequency-domain analysis of the accident time-series in PPC (for 1993–2009) and GCCs (for 1999–2007), prove that they are characterized by the existence of a periodic factor which (a) constitutes a permanent feature for the dynamic behavior of PPC's and GCCs' OHSS (occupational health and safety system), and (b) could be taken into account by risk managers in risk assessment, i.e., immediate suppressive measures must be taken place to abolish the danger source which is originated from the quasi-periodic appearance of the most important hazard sources

  17. Accident scenario diagnostics with neural networks

    International Nuclear Information System (INIS)

    Nuclear power plants are very complex systems. The diagnoses of transients or accident conditions is very difficult because a large amount of information, which is often noisy, or intermittent, or even incomplete, need to be processed in real time. To demonstrate their potential application to nuclear power plants, neural networks axe used to monitor the accident scenarios simulated by the training simulator of TVA's Watts Bar Nuclear Power Plant. A self-organization network is used to compress original data to reduce the total number of training patterns. Different accident scenarios are closely related to different key parameters which distinguish one accident scenario from another. Therefore, the accident scenarios can be monitored by a set of small size neural networks, called modular networks, each one of which monitors only one assigned accident scenario, to obtain fast training and recall. Sensitivity analysis is applied to select proper input variables for modular networks

  18. Severe Accident Research Program plan update

    International Nuclear Information System (INIS)

    In August 1989, the staff published NUREG-1365, ''Revised Severe Accident Research Program Plan.'' Since 1989, significant progress has been made in severe accident research to warrant an update to NUREG-1365. The staff has prepared this SARP Plan Update to: (1) Identify those issues that have been closed or are near completion, (2) Describe the progress in our understanding of important severe accident phenomena, (3) Define the long-term research that is directed at improving our understanding of severe accident phenomena and developing improved methods for assessing core melt progression, direct containment heating, and fuel-coolant interactions, and (4) Reflect the growing emphasis in two additional areas--advanced light water reactors, and support for the assessment of criteria for containment performance during severe accidents. The report describes recent major accomplishments in understanding the underlying phenomena that can occur during a severe accident. These include Mark I liner failure, severe accident scaling methodology, source term issues, core-concrete interactions, hydrogen transport and combustion, TMI-2 Vessel Investigation Project, and direct containment heating. The report also describes the major planned activities under the SARP over the next several years. These activities will focus on two phenomenological issues (core melt progression, and fuel-coolant interactions and debris coolability) that have significant uncertainties that impact our understanding and ability to predict severe accident phenomena and their effect on containment performance SARP will also focus on severe accident code development, assessment and validation. As the staff completes the research on severe accident issues that relate to current generation reactors, continued research will focus on efforts to independently evaluate the capability of new advanced light water reactor designs to withstand severe accidents

  19. Re-visiting the Piper Alpha accident

    OpenAIRE

    Dykesteen, Mette Kahrs

    2013-01-01

    The main objective of this thesis has been to re-visit the Piper Alpha accident using the latest version of the FLACS simulation code. In 1988/89 simulations of the gas explosion in the C Module of Piper Alpha were performed by Jan Roar Bakke and Idar Storvik at Christian Michelsen Institute, in conjunction with the investigation after the accident. For these simulations the computer code FLACS was used [1, 2]. In this thesis, the same simulation cases have been looked into, and the results o...

  20. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

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

  1. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-09-01

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

  2. Concealed holographic coding for security applications by using a moire technique

    DEFF Research Database (Denmark)

    Zhang, Xiangsu; Dalsgaard, Erik

    1997-01-01

    We present an optical coding technique that enhances the anticounterfeiting power of security holograms. The principles of the technique is based on the moire phenomenon. The code in the hologram has a phase pattern that is invisible and cannot be detected by optical equipment, so that imitation...... is extremely difficult. Holographic, photographic and embossing technique are used in fabricating coded holograms and decoders....

  3. Application of the Monte Carlo code DETEFF to efficiency calibrations for in situ gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Carrazana Gonzalez, J.; Cornejo Diaz, N. [Centre for Radiological Protection and Hygiene, P.O. Box 6195, Habana (Cuba); Jurado Vargas, M., E-mail: mjv@unex.es [Departamento de Fisica, Universidad de Extremadura, 06071 Badajoz (Spain)

    2012-05-15

    We studied the applicability of the Monte Carlo code DETEFF for the efficiency calibration of detectors for in situ gamma-ray spectrometry determinations of ground deposition activity levels. For this purpose, the code DETEFF was applied to a study case, and the calculated {sup 137}Cs activity deposition levels at four sites were compared with published values obtained both by soil sampling and by in situ measurements. The {sup 137}Cs ground deposition levels obtained with DETEFF were found to be equivalent to the results of the study case within the uncertainties involved. The code DETEFF could thus be used for the efficiency calibration of in situ gamma-ray spectrometry for the determination of ground deposition activity using the uniform slab model. It has the advantage of requiring far less simulation time than general Monte Carlo codes adapted for efficiency computation, which is essential for in situ gamma-ray spectrometry where the measurement configuration yields low detection efficiency. - Highlights: Black-Right-Pointing-Pointer Application of the code DETEFF to in situ gamma-ray spectrometry. Black-Right-Pointing-Pointer {sup 137}Cs ground deposition levels evaluated assuming a uniform slab model. Black-Right-Pointing-Pointer Code DETEFF allows a rapid efficiency calibration.

  4. Study of application of protective measures for the public and remediation of contaminated areas in case of nuclear and / or radiological accidents in Brazil; Estudo da aplicacao de medidas de protecao para o publico e de remediacao de areas contaminadas em caso de acidentes nucleares e/ou radiologicos no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Diogo Neves Gomes da

    2011-07-01

    Since the radiological accident in Goiania in 1987, the IRD (Institute of Radiological Protection and Dosimetry - IRD / CNEN) has been developing tools to support decision-making processes after a nuclear or radiological accident which leads to an environmental contamination and to an exposure of individuals the public These processes include the establishment of a supporting multicriteria model, which involves the application of protective and remediation measures of contaminated areas in tropical environments. In this study, it was performed an evaluation of the efficiency of these measures in order to determine the consequences of their implementation, based on results obtained from the code SIEM (Emergency Integrated System), which constitutes an environmental mode1 developed at IRD to simulate this type of accident. In order to perform this evaluation, it was first developed a database containing descriptions of various protection/remediation measures, which could be applied nationwide. Afterwards, some basic scenarios were established, considering the environmental, housing and food characteristics of the population of the vicinity of the nuclear power plants in Angra dos Reis (state of Rio de Janeiro). Thus, the accident simulations were carried out separately containing releases of {sup 137}Cs, {sup 90}Sr and 131 I. The results showed that the dose reduction varies according to the extent and the timing of the remediation measure applied. Although it is possible to establish some basic guidelines, generic solutions are not recommended, since the resulting doses are highly dependent on the actual situation. Any decision-making process should be made case by case, according to the actual conditions of the affected area and to the occupation characteristics and use of the affected areas, considering the characteristics of the source term of contamination, the time of the year in which the accident occurs, the local agricultural practices and food habits of

  5. APPLICATION OF INTEGER CODING ACCELERATING GENETIC ALGORITHM IN RECTANGULAR CUTTING STOCK PROBLEM

    Institute of Scientific and Technical Information of China (English)

    FANG Hui; YIN Guofu; LI Haiqing; PENG Biyou

    2006-01-01

    An improved genetic algorithm and its application to resolve cutting stock problem are presented. It is common to apply simple genetic algorithm (SGA) to cutting stock problem, but the huge amount of computing of SGA is a serious problem in practical application. Accelerating genetic algorithm (AGA) based on integer coding and AGA's detailed steps are developed to reduce the amount of computation, and a new kind of rectangular parts blank layout algorithm is designed for rectangular cutting stock problem. SGA is adopted to produce individuals within given evolution process, and the variation interval of these individuals is taken as initial domain of the next optimization process, thus shrinks searching range intensively and accelerates the evaluation process of SGA.To enhance the diversity of population and to avoid the algorithm stagnates at local optimization result, fixed number of individuals are produced randomly and replace the same number of parents in every evaluation process. According to the computational experiment, it is observed that this improved GA converges much sooner than SGA, and is able to get the balance of good result and high efficiency in the process of optimization for rectangular cutting stock problem.

  6. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  7. Applicability of the three-dimensional transport code Tort to the shielding analysis of the prototype FBR Monju

    International Nuclear Information System (INIS)

    Shielding design of Monju was performed in 1980's by using the two-dimensional discrete ordinates transport code, DOT3.5. In view of complexity of the three-dimensional shielding geometry of Monju, the three-dimensional discrete ordinates transport code, TORT(2), has been applied to shielding measurement analyses of Monju in an attempt to prove practical usefulness of the code and to learn how much margin is associated with the shielding design performed by DOT3.5. This study has confirmed that TORT can practically be applied to the shielding measurement analyses of Monju, and has provided significant improvement in calculation accuracy thanks to its three-dimensional geometry employed, making the code applicable to the Monju shielding design evaluation analyses together with pre- and post-analyses of the shielding measurement now being planned. (authors)

  8. Applicability of the three-dimensional transport code Tort to the shielding analysis of the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Takako, Shiraki [Mitsubishi Heavy Industries, Ltd (Japan); Shin, Usami; Zenro, Suzuoki; Takehide, Deshimaru [Japan Nuclear Cycle Development Institute (Japan); Kenji, Sasaki; Keiko, Tada; Hitoshi, Yokobori [Advanced Reactor Technology Co., Ltd (Japan)

    2003-07-01

    Shielding design of Monju was performed in 1980's by using the two-dimensional discrete ordinates transport code, DOT3.5. In view of complexity of the three-dimensional shielding geometry of Monju, the three-dimensional discrete ordinates transport code, TORT(2), has been applied to shielding measurement analyses of Monju in an attempt to prove practical usefulness of the code and to learn how much margin is associated with the shielding design performed by DOT3.5. This study has confirmed that TORT can practically be applied to the shielding measurement analyses of Monju, and has provided significant improvement in calculation accuracy thanks to its three-dimensional geometry employed, making the code applicable to the Monju shielding design evaluation analyses together with pre- and post-analyses of the shielding measurement now being planned. (authors)

  9. Application of the Monte Carlo code DETEFF to efficiency calibrations for in situ gamma-ray spectrometry.

    Science.gov (United States)

    Carrazana González, J; Cornejo Díaz, N; Jurado Vargas, M

    2012-05-01

    We studied the applicability of the Monte Carlo code DETEFF for the efficiency calibration of detectors for in situ gamma-ray spectrometry determinations of ground deposition activity levels. For this purpose, the code DETEFF was applied to a study case, and the calculated (137)Cs activity deposition levels at four sites were compared with published values obtained both by soil sampling and by in situ measurements. The (137)Cs ground deposition levels obtained with DETEFF were found to be equivalent to the results of the study case within the uncertainties involved. The code DETEFF could thus be used for the efficiency calibration of in situ gamma-ray spectrometry for the determination of ground deposition activity using the uniform slab model. It has the advantage of requiring far less simulation time than general Monte Carlo codes adapted for efficiency computation, which is essential for in situ gamma-ray spectrometry where the measurement configuration yields low detection efficiency. PMID:22336296

  10. a Simplified Methodology for the Prediction of the Small Break Loss-Of Accident.

    Science.gov (United States)

    Ward, Leonard William

    1988-12-01

    This thesis describes a complete methodology which has allowed for the development of a faster than real time computer program designed to simulate a small break loss -of-coolant accident in the primary system of a pressurized water reactor. By developing an understanding of the major phenomenon governing the small break LOCA fluid response, the system model representation can be greatly simplified leading to a very fast executing transient system blowdown code. Because of the fast execution times, the CULSETS code, or Columbia University Loss-of-Coolant Accident and System Excursion Transient Simulator code, is ideal for performing parametric studies of Emergency Core Cooling system or assessing the consequences of the many operator actions performed to place the system in a long term cooling mode following a small break LOCA. While the methodology was designed with specific application to the small break loss-of-coolant accident, it can also be used to simulate loss-of-feedwater, steam line breaks, and steam generator tube rupture events. The code is easily adaptable to a personal computer and could also be modified to provide the primary and secondary system responses to supply the required inputs to a simulator for a pressurized water reactor.

  11. Development of a fast running multidimensional thermal-hydraulic code to be readily coupled with multidimensional neutronic tools, applicable to modular high temperature reactors

    International Nuclear Information System (INIS)

    Modular High Temperature Reactors (HTRs) are considered as one of the most promising next generation reactors which will fulfill the future energy demand. The inherent safety is the most attractive feature of this type of reactor along with simplicity in design, operation and maintenance. Since the reactor is safe during any accident conditions without the actuation of any external safety systems, it is considered to be a inherently safe reactor. With its offered inherent safety features, the reactor responses solely from the reactor's physical properties, hence any dangerous situation will be avoided. The inherent safety feature of this reactor depends entirely on the correct design of this reactor. The power density in the core, radius and height of the core, properties of the materials used and its configuration must be chosen in such a way that the decay heat produced in the core during any accident can be released to the surrounding by natural heat transfer phenomena without any help of external safety features. In addition, possible reactivity insertions into the core are limited such that the corresponding temperature increases of the fuels stay always below the fuel's temperature design limit. Along with its inherent safety feature, the reactor must be designed such a way that it offers a competitive economics. The objective of this endeavor is to develop a fast running/multidimensional code which can be used to analyze, design and safety related issues in modular high temperature reactors. The program shall be generally applicable for modular HTRs (e.g pebble fuel, block fuel elements). Operational conditions with forced cooling as well as accident situations with heat removal by conduction and natural circulation shall be covered. Coupling to a reactor physics code shall be provided to account for the feedback of neutronics and thermal-hydraulics. Emphasis is on capturing essential effects resulting from three-dimensional features (e.g. single

  12. 10 CFR 50.67 - Accident source term.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The... to January 10, 1997, who seek to revise the current accident source term used in their design...

  13. Analysis of the metallic containment integrity of Angra 2/3 reactor under the effects of the design basis accident

    International Nuclear Information System (INIS)

    The application of Condru 4 computer code, developed to determine the maximum values of pressure and temperature that occur inside the metallic containment building of PWR nuclear power plants, in case of a hypothetic accident - LOCA - considered as a Design Basic Accident - DBA. The hypothesis, input and results for the simulation of a loss of coolant in the hot leg of the Angra-2/3 reactors, considered as the most critical case for that Kind of project, are presented. The analysis was made with input provided by the manufacturer. (Author)

  14. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  15. TITAN: a computer program for accident occurrence frequency analyses by component Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kanai, Shigeru [Fuji Research Institute Corporation, Tokyo (Japan)

    2000-04-01

    In a plant system consisting of complex equipments and components for a reprocessing facility, there might be grace time between an initiating event and a resultant serious accident, allowing operating personnel to take remedial actions, thus, terminating the ongoing accident sequence. A component Monte Carlo simulation computer program TITAN has been developed to analyze such a complex reliability model including the grace time without any difficulty to obtain an accident occurrence frequency. Firstly, basic methods for the component Monte Carlo simulation is introduced to obtain an accident occurrence frequency, and then, the basic performance such as precision, convergence, and parallelization of calculation, is shown through calculation of a prototype accident sequence model. As an example to illustrate applicability to a real scale plant model, a red oil explosion in a German reprocessing plant model is simulated to show that TITAN can give an accident occurrence frequency with relatively good accuracy. Moreover, results of uncertainty analyses by TITAN are rendered to show another performance, and a proposal is made for introducing of a new input-data format to adapt the component Monte Carlo simulation. The present paper describes the calculational method, performance, applicability to a real scale, and new proposal for the TITAN code. In the Appendixes, a conventional analytical method is shown to avoid complex and laborious calculation to obtain a strict solution of accident occurrence frequency, compared with Monte Carlo method. The user's manual and the list/structure of program are also contained in the Appendixes to facilitate TITAN computer program usage. (author)

  16. Function and Application Areas in Medicine of Non-Coding RNA

    Directory of Open Access Journals (Sweden)

    Figen Guzelgul

    2009-06-01

    Full Text Available RNA is the genetic material converting the genetic code that it gets from DNA into protein. While less than 2 % of RNA is converted into protein , more than 98 % of it can not be converted into protein and named as non-coding RNAs. 70 % of noncoding RNAs consists of introns , however, the rest part of them consists of exons. Non-coding RNAs are examined in two classes according to their size and functions. Whereas they are classified as long non-coding and small non-coding RNAs according to their size , they are grouped as housekeeping non-coding RNAs and regulating non-coding RNAs according to their function. For long years ,these non-coding RNAs have been considered as non-functional. However, today, it has been proved that these non-coding RNAs play role in regulating genes and in structural, functional and catalitic roles of RNAs converted into protein. Due to its taking a role in gene silencing mechanism, particularly in medical world , non-coding RNAs have led to significant developments. RNAi technolgy , which is used in designing drugs to be used in treatment of various diseases , is a ray of hope for medical world. [Archives Medical Review Journal 2009; 18(3.000: 141-155

  17. Optimal binary phase codes and sidelobe-free decoding filters with application to incoherent scatter radar

    Directory of Open Access Journals (Sweden)

    M. S. Lehtinen

    2004-04-01

    Full Text Available This paper presents binary phase codes and corresponding decoding filters which are optimal in the sense that they produce no sidelobes and they maximise the signal-to-noise ratio (SNR henceforth. The search is made by investigating all possible binary phase codes with a given length. After selecting the code, the first step is to find a filter which produces no sidelobes. This is possible for all codes with no zeros in the frequency domain, and it turns out that most codes satisfy this requirement. An example of a code which cannot be decoded in this way is a code with a single phase, i.e. a long pulse. The second step is to investigate the SNR performance of the codes. Then the optimal code of a given length is the one with the highest SNR at the filter output. All codes with lengths of 3–25 bits were studied, which means investigating 33554428 binary phase codes. It turns out that all Barker codes except the 11-bit code are optimal in the above sense. It is well known that the performance of matched-filter decoding of Barker codes is better than decoding without sidelobes. In the case of the 7-bit Barker code, it is shown here that the SNR given by sidelobe-free decoding is nearly 30% worse than that of standard decoding, but for the 13-bit code sidelobe-free decoding is only about 5% worse. The deterioration of SNR should be evaluated against the benefits gained in disposing of the sidelobes, which, even for the 13-bit code, contribute by 7.1% to the total signal power from a homogeneous target. Thus, regions of weak scattering can be contaminated by the sidelobes from neighbouring layers of strong scattering, causing broadening of thin spatial structures and giving a lower spatial resolution than implied by the bit length. A practical example is shown where sidelobes mask a weak signal when the standard matched filter is used in the analysis. An improvement is achieved when sidelobe-free filtering is carried out.

    Key words. Radio

  18. Bridging Inter-flow and Intra-flow Network Coding for Video Applications

    DEFF Research Database (Denmark)

    Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani;

    2013-01-01

    Network Coding (NC) for Wireless Mesh Network (WMN) has received a lot of attention from the research community in recent years due to its ability to provide higher throughput, reliability, and efficiency in the use of the available wireless spectrum. NC has had two main research lines focused...... on inter-flow and intra-flow coding. However, these two domains have historically been studied separately. This paper proposes a testbed implementation of CORE, the first protocol to bridge both approaches such that the structure of intra-flow coding can be exploited during inter-flow coding to both...... enhance reliability, common of the former, while maintaining an efficient spectrum usage, typical of the latter. This paper uses the intuition provided in [1] to propose a practical implementation of the protocol leveraging Random Linear Network Coding (RLNC) for intra-flow coding, a credit based packet...

  19. The Hybrid Detailed / Statistical Opacity Code SCO-RCG: New Developments and Applications

    OpenAIRE

    Pain, Jean-Christophe; Gilleron, Franck; Porcherot, Quentin; Blenski, Thomas

    2013-01-01

    We present the hybrid opacity code SCO-RCG which combines statistical approaches with fine-structure calculations. Radial integrals needed for the computation of detailed transition arrays are calculated by the code SCO (Super-configuration Code for Opacity), which calculates atomic structure at finite temperature and density, taking into account plasma effects on the wave-functions. Levels and spectral lines are then computed by an adapted RCG routine of R. D. Cowan. SCO-RCG now includes the...

  20. Severe Accident Recriticality Analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

    1999-11-01

    Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B{sub 4}C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

  1. Application of computational fluid dynamics methods to improve thermal hydraulic code analysis

    Science.gov (United States)

    Sentell, Dennis Shannon, Jr.

    A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative evaluation of the experimental data, the thermal hydraulic code results and the computational fluid dynamics code results provides an opportunity to validate the best-estimate thermal hydraulic code's treatment of a natural circulation loop and provide insights into expanded use of the computational fluid dynamics code in future designs and operations. Additionally, a sensitivity analysis is conducted to determine those physical phenomena most impactful on operations of the proposed reactor's natural circulation loop. The combination of the comparative evaluation and sensitivity analysis provides the resources for increased confidence in model developments for natural circulation loops and provides for reliability improvements of the thermal hydraulic code.

  2. Development and application of a multi-fluid simulation code for modeling interpenetrating plasmas

    Science.gov (United States)

    Khodak, M.; Berger, R. L.; Chapman, T.; Hittinger, J. A. F.

    2015-11-01

    A multi-fluid model, with independent velocities for all species, is developed and implemented for the numerical simulation of the interpenetration of colliding plasmas. The Euler equations for fluid flow, coupled through electron-ion and ion-ion collisional drag terms, thermal equilibration terms, and the electric field, are solved for each ion species with the electrons treated under a quasineutrality assumption. Fourth-order spatial convergence in smooth regions is achieved using flux-conservative iterative time integration and a Weighted Essentially Non-Oscillatory (WENO) finite volume scheme employing an approximate Riemann solver. Analytic solutions of well-known shock tube tests and spectral solutions of the linearized coupled system are used to test the implementation, and the model is further numerically compared to interpenetration experiments such as those of J.S. Ross et al. [Phys. Rev. Lett. 110 145005 (2013)]. This work has applications to laser-plasma interactions, specifically to hohlraum physics, as well as to modeling laboratory experiments of collisionless shocks important in astrophysical plasmas. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract No. DE-AC52-07NA27344 and funded by the Laboratory Research and Development Program at LLNL under project code 15-ERD-038.

  3. Image embedded coding with edge preservation based on local variance analysis for mobile applications

    Science.gov (United States)

    Luo, Gaoyong; Osypiw, David

    2006-02-01

    Transmitting digital images via mobile device is often subject to bandwidth which are incompatible with high data rates. Embedded coding for progressive image transmission has recently gained popularity in image compression community. However, current progressive wavelet-based image coders tend to send information on the lowest-frequency wavelet coefficients first. At very low bit rates, images compressed are therefore dominated by low frequency information, where high frequency components belonging to edges are lost leading to blurring the signal features. This paper presents a new image coder employing edge preservation based on local variance analysis to improve the visual appearance and recognizability of compressed images. The analysis and compression is performed by dividing an image into blocks. Fast lifting wavelet transform is developed with the advantages of being computationally efficient and boundary effects minimized by changing wavelet shape for handling filtering near the boundaries. A modified SPIHT algorithm with more bits used to encode the wavelet coefficients and transmitting fewer bits in the sorting pass for performance improvement, is implemented to reduce the correlation of the coefficients at scalable bit rates. Local variance estimation and edge strength measurement can effectively determine the best bit allocation for each block to preserve the local features by assigning more bits for blocks containing more edges with higher variance and edge strength. Experimental results demonstrate that the method performs well both visually and in terms of MSE and PSNR. The proposed image coder provides a potential solution with parallel computation and less memory requirements for mobile applications.

  4. A portable platform for accelerated PIC codes and its application to GPUs using OpenACC

    Science.gov (United States)

    Hariri, F.; Tran, T. M.; Jocksch, A.; Lanti, E.; Progsch, J.; Messmer, P.; Brunner, S.; Gheller, C.; Villard, L.

    2016-10-01

    We present a portable platform, called PIC_ENGINE, for accelerating Particle-In-Cell (PIC) codes on heterogeneous many-core architectures such as Graphic Processing Units (GPUs). The aim of this development is efficient simulations on future exascale systems by allowing different parallelization strategies depending on the application problem and the specific architecture. To this end, this platform contains the basic steps of the PIC algorithm and has been designed as a test bed for different algorithmic options and data structures. Among the architectures that this engine can explore, particular attention is given here to systems equipped with GPUs. The study demonstrates that our portable PIC implementation based on the OpenACC programming model can achieve performance closely matching theoretical predictions. Using the Cray XC30 system, Piz Daint, at the Swiss National Supercomputing Centre (CSCS), we show that PIC_ENGINE running on an NVIDIA Kepler K20X GPU can outperform the one on an Intel Sandy bridge 8-core CPU by a factor of 3.4.

  5. Application to MISR Land Products of an RPV Model Inversion Package Using Adjoint and Hessian Codes

    Science.gov (United States)

    Lavergne, T.; Kaminski, T.; Pinty, B.; Taberner, M.; Gobron, N.; Verstraete, M. M.; Vossbeck, M.; Widlowski, J.-L.; Giering, R.

    The capability of the non-linear Rahman-Pinty-Verstraete RPV model to 1 accurately fit a large variety of Bidirectional Reflectance Factor BRF fields and 2 return parameter values of interest for land surface applications motivate the development of a computer efficient inversion package The present paper describes such a package based on the 3 and 4 parameter versions of the RPV model This software environment implements the adjoint code generated using automatic differentiation techniques of the cost function This cost function itself balances two main contributions reflecting 1 the a priori knowledge on the model parameter values and 2 BRF uncertainties together with the requirement to minimize the mismatch between the measurements and the RPV simulations The individual weights of these contributions are specified notably via covariance matrices of the uncertainties in the a priori knowledge on the model parameters and the observations This package also reports on the probability density functions of the retrieved model parameter values that thus permit the user to evaluate the a posteriori uncertainties on these retrievals This is achieved by evaluating the Hessian of the cost function at its minimum Results from a variety of tests are shown in order to document and analyze software performance against complex synthetic BRF fields simulated by radiation transfer models as well as against actual MISR-derived surface BRF products

  6. Construction of FuzzyFind Dictionary using Golay Coding Transformation for Searching Applications

    Directory of Open Access Journals (Sweden)

    Kamran Kowsari

    2015-03-01

    Full Text Available searching through a large volume of data is very critical for companies, scientists, and searching engines applications due to time complexity and memory complexity. In this paper, a new technique of generating FuzzyFind Dictionary for text mining was introduced. We simply mapped the 23 bits of the English alphabet into a FuzzyFind Dictionary or more than 23 bits by using more FuzzyFind Dictionary, and reflecting the presence or absence of particular letters. This representation preserves closeness of word distortions in terms of closeness of the created binary vectors within Hamming distance of 2 deviations. This paper talks about the Golay Coding Transformation Hash Table and how it can be used on a FuzzyFind Dictionary as a new technology for using in searching through big data. This method is introduced by linear time complexity for generating the dictionary and constant time complexity to access the data and update by new data sets, also updating for new data sets is linear time depends on new data points. This technique is based on searching only for letters of English that each segment has 23 bits, and also we have more than 23-bit and also it could work with more segments as reference table.

  7. Application of STAV5 code for the analysis of fission gas release in power reactor rods

    International Nuclear Information System (INIS)

    STAV5 is a design code for calculation of temperatures, fission gas release and rod pressure in BWR and PWR fuel rods. It includes the treatment of pellet cracks affecting conductivity and thermal expansion, gap closure by eccentric or relocated pellet fragments and oxide and crud build-up on the clad outer surface. The fission gas release model consists of two parts: High temperature release based on grain boundary saturation and low temperature release varying with fission rate of different isotopes. STAV5 has been benchmarked with a number of inpile thermal measurement experiments to as high burnup as 25 MWd/kg U. The main application of STAV5 is as a routine design tool for power reactor rods. It is also used to compare with PIE data. Examples are given from the analyses of fission gas release data from BWR rods from Oskarshamn 1 and Barsebeck 1 as well as PWR rods from Maine Yankee initial cores. The STAV5 evaluation show the importance of power histories, densification and the position in the assembly. (author)

  8. Construction of FuzzyFind Dictionary using Golay Coding Transformation for Searching Applications

    Science.gov (United States)

    Kowsari, Kamram

    2015-03-01

    searching through a large volume of data is very critical for companies, scientists, and searching engines applications due to time complexity and memory complexity. In this paper, a new technique of generating FuzzyFind Dictionary for text mining was introduced. We simply mapped the 23 bits of the English alphabet into a FuzzyFind Dictionary or more than 23 bits by using more FuzzyFind Dictionary, and reflecting the presence or absence of particular letters. This representation preserves closeness of word distortions in terms of closeness of the created binary vectors within Hamming distance of 2 deviations. This paper talks about the Golay Coding Transformation Hash Table and how it can be used on a FuzzyFind Dictionary as a new technology for using in searching through big data. This method is introduced by linear time complexity for generating the dictionary and constant time complexity to access the data and update by new data sets, also updating for new data sets is linear time depends on new data points. This technique is based on searching only for letters of English that each segment has 23 bits, and also we have more than 23-bit and also it could work with more segments as reference table.

  9. Fibre segment interferometry using code-division multiplexed optical signal processing for strain sensing applications

    International Nuclear Information System (INIS)

    A novel optical signal processing scheme for multiplexing fibre segment interferometers is proposed. The continuous-wave, homodyne technique combines code-division multiplexing with single-sideband modulation. It uses only one electro-optic phase modulator to achieve both range separation and quadrature interferometric phase measurement. This scheme is applied to fibre segment interferometry, where a number of long-gauge length interferometric fibre sensors are formed by subtracting pairs of signals from equidistantly placed, weak back reflectors. In this work we give a detailed account of the signal processing involved and, in particular, explore aspects such as electronic bandwidth requirements, noise, crosstalk and linearity, which are important design considerations. A signal bandwidth of ±20 kHz permits the resolution of phase change rates of 2.5 × 104 rad s−1 for each of the four 16.5 m long segments in our setup. We show that dynamic strain resolutions below 0.2 nanostrain Hz−0.5 at 2 m sensor gauge length are achievable, even with an inexpensive diode laser. When used in applications that require only relative strain change measurements, this scheme compares well to more established techniques and can provide high-fidelity yet cost-effective measurements. (paper)

  10. Revised Safety Code A2

    CERN Document Server

    SC Secretariat

    2005-01-01

    Please note that the revised Safety Code A2 (Code A2 rev.) entitled "REPORTING OF ACCIDENTS AND NEAR MISSES" is available on the web at the following url: https://edms.cern.ch/document/335502/LAST_RELEASED Paper copies can also be obtained from the SC Unit Secretariat, e-mail: sc.secretariat@cern.ch SC Secretariat

  11. Application of QR code in mobile library%QR码在手机图书馆中的应用

    Institute of Scientific and Technical Information of China (English)

    张蓓; 张成昱; 窦天芳; 远红亮

    2012-01-01

      As one type of two-dimensional code, QR code has the advantages of large capacity and high reliability. With the construction of the mobile library and the combination with SMS and mobile websites, QR code is widely used in scenarios like OPAC and study room reservations. QR code has extended the library services in time, space and manner. This paper briefly introduces the basic situation of QR code and analyzes several applications of QR code in the mobile library. It also points out some problems which need attention while promoting QR code in libraries. With the arrival of 3G era and popularization of smartphones, QR code will be given full play in library services.%  QR码作为二维码中的一种,具有容量大、可靠性高等优势。伴随手机图书馆的逐步建设,QR码与短信、手机网站等结合,广泛应用于书目联机查询系统(OPAC)、研读间预约等场景,促使图书馆服务的时空、方式得到拓展和延伸。简要介绍QR码的基本情况后,针对目前手机图书馆中QR码的几种具体应用形式进行分析,并指出该技术在图书馆推广过程中需要注意的问题。随着3G时代的到来、智能手机的普及,QR码在图书馆服务中将发挥更大的作用

  12. Application of the micronucleus assay performed by different scorers in case of large-scale radiation accidents

    Directory of Open Access Journals (Sweden)

    Rawojć Kamila

    2015-09-01

    Full Text Available Mass casualty scenarios of radiation exposure require high throughput biological dosimetry techniques for population triage, in order to rapidly identify individuals, who require clinical treatment. Accurate dose estimates can be made by biological dosimetry, to predict the acute radiation syndrome (ARS within days after a radiation accident or a malicious act involving radiation. Timely information on dose is important for the medical management of acutely irradiated persons [1]. The aim of the study was to evaluate the usefulness of the micronuclei (MNi scoring procedure in an experimental mode, where 500 binucleated cells were analyzed in different exposure dose ranges. Whole-body exposure was simulated in an in vitro experiment by irradiating whole blood collected from one healthy donor with 60 MeV protons and 250 keV X-rays, in the dose range of 0.3-4.0 Gy. For achieving meaningful results, sample scoring was performed by three independent persons, who followed guidelines described in detail by Fenech et al. [2, 3]. Compared results revealed no significant differences between scorers, which has important meaning in reducing the analysis time. Moreover, presented data based on 500 cells distribution, show that there are significant differences between MNi yields after 1.0 Gy exposure of blood for both protons and X-rays, implicating this experimental mode as appropriate for the distinction between high and low dose-exposed individuals, which allows early classification of exposed victims into clinically relevant subgroups.

  13. Research and Application of Auxiliary Optimization Technology of Power Grid Accident Processing Based on the Mode of Regulation and Control Integration

    Directory of Open Access Journals (Sweden)

    Cui Houzhen

    2015-01-01

    Full Text Available Accident processing is the most important link of the scheduling of daily monitoring. The improvement of intelligent level is of great significance for improving the efficiency of accident processing scheduling, shortening the time of accident processing and preventing further deterioration of accidents. According to features of accident processing scheduling, this paper puts forward an integrated framework of aid decision-making of online accident processing based on large power grid, and carries out a study from five aspects, namely integrated information support platform, risk perception in advance, online fault diagnosis, aid decision-making afterwards and visual display, so as to conduct real-time tracking on operating state of power grid, eliminate potential safety hazards of power grid and upgrade power grid from “manual analysis” scheduling to “intelligent analysis” scheduling.

  14. Supervisor's accident investigation handbook

    International Nuclear Information System (INIS)

    This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

  15. Simulation and study on the γ response spectrum of BGO detector by the application of monte carlo code MOCA

    International Nuclear Information System (INIS)

    Application of Monte Carlo method to build spectra library is useful to reduce experiment workload in Prompt Gamma Neutron Activation Analysis (PGNAA). The new Monte Carlo Code MOCA was used to simulate the response spectra of BGO detector for gamma rays from 137Cs, 60Co and neutron induced gamma rays from S and Ti. The results were compared with general code MCNP, show that the agreement of MOCA between simulation and experiment is better than MCNP. This research indicates that building spectra library by Monte Carlo method is feasible. (authors)

  16. Meta-code for systematic analysis of chemical addition (SACHA): application to fluorination of C70 and carbon nanostructure growth.

    Science.gov (United States)

    Ewels, Christopher P; Lier, Gregory Van; Geerlings, Paul; Charlier, Jean-Christophe

    2007-01-01

    We present a new computer program able to systematically study chemical addition to and growth or evolution of carbon nanostructures. SACHA is a meta-code able to exploit a wide variety of pre-existing molecular structure codes, automating the otherwise onerous task of constructing, running, and analyzing the large number of input files that are required when exploring structural isomers and addition paths. By way of examples we consider fluorination of the fullerene cage C70 and carbon nanostructure growth through C2 addition. We discuss the possibilities for extension of this technique to rapidly and efficiently explore structural energy landscapes and application to other areas of chemical and materials research.

  17. Estimation of Dynamic Discrete Games Using the Nested Pseudo Likelihood Algorithm: Code and Application

    OpenAIRE

    Aguirregabiria, Victor

    2009-01-01

    This document describes program code for the solution and estimation of dynamic discrete games of incomplete information using the Nested Pseudo Likelihood (NPL) method in Aguirregabiria and Mira (2007). The code is illustrated using a dynamic game of store location by retail chains, and actual data from McDonalds and Burger King.

  18. Applications of the lahet simulation code to relativistic heavy ion detectors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, L.; Gavron, A. [Los Alamos National Lab., NM (United States)

    1991-12-31

    The Los Alamos High Energy Transport (LAHET) simulation code has been applied to test beam data from the lead/scintillator Participant Calorimeter of BNL AGS experiment E814. The LAHET code treats hadronic interactions with the LANL version of the Oak Ridge code HETC. LAHET has now been expanded to handle hadrons with kinetic energies greater than 5 GeV with the FLUKA code, while HETC is used exclusively below 2.0 GeV. FLUKA is phased in linearly between 2.0 and 5.0 GeV. Transport of electrons and photons is done with EGS4, and an interface to the Los Alamos HMCNP3B library based code is provided to analyze neutrons with kinetic energies less than 20 MeV. Excellent agreement is found between the test data and simulation, and results for 2.46 GeV/c protons and pions are illustrated in this article.

  19. Development and Application of a Plant Code to the Analysis of Transients in Integrated Reactors

    International Nuclear Information System (INIS)

    In this work, a secondary system model for a CAREM-25 type nuclear power plant was developed.A two-phase flow homogenous model was used and found adequate for the scope of the present work.A finite difference scheme was used for the numerical implementation of the model.This model was coupled to the HUARPE code, a primary circuit code, in order to obtain a plant code.This plant code was used to analyze the inherent response of the system, without control feedback loops, for a transient of steam generator feed-water mass flow reduction.The results obtained are satisfactory, but a validation against other plant codes is still necessary

  20. Application of the ballooning analysis code MATARE on a generic PWR fuel assembly

    International Nuclear Information System (INIS)

    The MATARE (MAbel-TAlink-RElap) code is a new multi-pin deformation analysis code created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. A multi-pin representation of different zones of a typical PWR fuel assembly under post-LOCA reflooding conditions was analysed including some of the most relevant features that characterise a typical nuclear reactor fuel assembly and evaluate their effect on the behaviour of the fuel rods under conditions leading to clad ballooning. The code was able to simulate the deformation of wide regions of a fuel assembly under reflood conditions and has shown how differences in pin pressure and the presence of rod with burnable poisons and control rod guide thimbles also contribute to a substantial incoherent ballooning in agreement with the experimental data. (author)